Method of detecting failed fuels

International Nuclear Information System (INIS)

Ishizaki, Hideaki; Suzumura, Takeshi.

1982-01-01

Purpose: To enable the settlement of the temperature of an adequate filling high temperature pure water by detecting the outlet temperature of a high temperature pure water filling tube to a fuel assembly to control the heating of the pure water and detecting the failed fuel due to the sampling of the pure water. Method: A temperature sensor is provided at a water tube connected to a sipping cap for filling high temperature pure water to detect the temperature of the high temperature pure water at the outlet of the tube, and the temperature is confirmed by a temperature indicator. A heater is controlled on the basis of this confirmation, an adequate high temperature pure water is filled in the fuel assembly, and the pure water is replaced with coolant. Then, it is sampled to settle the adequate temperature of the high temperature coolant used for detecting the failure of the fuel assembly. As a result, the sipping effect does not decrease, and the failed fuel can be precisely detected. (Yoshihara, H.)

Development of failed fuel detection system for PWR (III)

International Nuclear Information System (INIS)

Hwang, Churl Kew; Kang, Hee Dong; Jeong, Seung Ho; Cho, Byung Sub; Yoon, Byeong Joo; Yoon, Jae Seong

1987-12-01

Ultrasonic transducers satisfying the conditions for failed fuel rod detection for failed fuel rod detection have been designed and built. And performance tests for them have been carried out. Ultrasonic signal processing units, a manipulator guiding the ultrasonic probe through the fuel assembly lanes and its control units have been constructed. The performance of the system has been verified experimentally to be successful in failed fuel rod detection. (Author)

Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

International Nuclear Information System (INIS)

Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

2010-01-01

A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

A method of failed fuel detection

International Nuclear Information System (INIS)

Uchida, Shunsuke; Utamura, Motoaki; Urata, Megumu.

1976-01-01

Object: To keep the coolant fed to a fuel assembly at a level below the temperature of existing coolant to detect a failed fuel with high accuracy without using a heater. Structure: When a coolant in a coolant pool disposed at the upper part of a reactor container is fed by a coolant feed system into a fuel assembly through a cap to fill therewith and exchange while forming a boundary layer between said coolant and the existing coolant, the temperature distribution of the feed coolant is heated by fuel rods so that the upper part is low whereas the lower part is high. Then, the lower coolant is upwardly moved by the agitating action and fission products leaked through a failed opening at the lower part of the fuel assembly and easily extracted by the sampling system. (Yoshino, Y.)

Failed fuel detection device

International Nuclear Information System (INIS)

Doi, Akira.

1994-01-01

The device of the present invention concerns a failed fuel detection device for a nuclear reactor, such as an FBR type reactor, using electroconductive coolants. A sampling port is disposed at the upper portion of the fuel assembly so as to cover the assembly, so that coolants in the fuel assembly are sampled to improve a device for detecting fuel failure. That is, when coolants in the fuel assembly are sampled from the sampling port, the flow of electroconductive coolants in an sampling tube is detected by a flowmeter, to control an electromagnetic pump. The flow of electroconductive coolants is stopped against the waterhead pressure and dynamic pressure of the conductive coolants, and a predetermined amount of the coolants is pumped up to the sampling tank. Gas is supplied to the pumped up coolants so that fissile products are transferred from the coolants to a gas phase. Radiation in the gas in a gas recycling system is measured to detect presence of fuel failure. (I.S.)

Experience with failed or damaged spent fuel and its impacts on handling

International Nuclear Information System (INIS)

Bailey, W.J.

1989-12-01

Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

Method for detecting a failed fuel

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke.

1976-01-01

Purpose: To provide a method for the detection of failed fuel by pouring hot water, in which pouring speed of liquid to be poured and temperature of the liquid are controlled to prevent the leakage of the liquid. Constitution: The method comprises blocking the top of a fuel assembly arranged in coolant to stop a flow of coolant, pouring a liquid higher in temperature than that of coolant into the fuel assembly, sampling the liquid poured, and measuring the concentration of radioactivity of coolant already subjected to sampling to detect a failed fuel. At this time, controlling is made so that the pouring speed of the poured liquid is set to about 25 l/min, and an increased portion of temperature from the temperature of liquid to the temperature of coolant is set to a level less than about 15 0 C. (Furukawa, Y.)

Reliability testing of failed fuel location system

International Nuclear Information System (INIS)

Vieru, G.

1996-01-01

This paper presents the experimental reliability tests performed in order to prove the reliability parameters for Failed Fuel Location System (FFLS), equipment used to detect in which channel of a particular heat transport loop a fuel failure is located, and to find in which channel what particular bundle pair is failed. To do so, D20 samples from each reactor channel are sequentially monitored to detect a comparatively high level of delayed neutron activity. 15 refs, 8 figs, 2 tabs

An integrated approach for investigation of failed nuclear fuel used at NPP Cernavoda Unit 1

International Nuclear Information System (INIS)

Tuturici, I.L.; Parvan, M.; Popov, M.; Dobrin, R.; Staicu, C.

1996-01-01

At NPP Cernavoda-Unit 1 the fuel surveillance and the defect detection system in operation are based on monitoring the coolant activity concentration and on measuring the flux of delayed neutrons emitted by some short-lived fission products. In order to identify the failed fuel underwater non-destructive examination has to be performed. The major interest for the availability of underwater examination consists in the necessity of a speedy acquisition of the data on failed fuel in operation and of appropriate follow-up actions to be taken. Often the identification operation will be followed by more detailed examinations on selected fuel rods in the hot cells of the Post-irradiation Examination Laboratory of the Institute for Nuclear Research at Pitesti. Transfer of selected fuel rods will be done by the use of a type B(U) road transportation cask. Such an integrated approach will help to keep the level of activity concentration of the primary circuit well below the authorized limits. (author). 2 figs., 1 tab., 2 refs

Failed fuel detection device

International Nuclear Information System (INIS)

Kawai, Masayoshi; Hayashida, Yoshihisa; Niidome, Jiro.

1985-01-01

Purpose: To prevent intrusion of background neutrons to neutron detectors thereby improve the S/N ratio of the detectors in the failed fuel detection device of LMFBR type reactors. Constitution: Neutrons from the reactor core pass through the gaps around the penetration holes in which the primary pipeways pass through the concrete shielding walls and pass through the gaps between the thermal shielding members and the neutron moderating shielding members of the failed fuel detection device and then intrude into the neutron detectors. In view of the above, inner neutron moderating shielding members and movable or resilient neutron shielding members are disposed to the inside of the neutron moderating shielding member. Graphite or carbon hydrides such as paraffin or synthetic resin with a large neutron moderation effect are used as the outer moderating shielding member and materials such as boron or carbon are used for the inner members. As a result, the background neutrons are shielded by the inner neutron moderating shielding members and the resilient neutron shielding members, by which the S/N ratio of the neutron detectors can be increased to 2 - 4 times. (Moriyama, K.)

Synthesis Report on the understanding of failed LMFBR fuel element performance

International Nuclear Information System (INIS)

Plitz, H.; Bagley, K.; Harbourne, B.

1990-07-01

In the coarse of LMFBR operation fuel element failures cannot entirely be avoided as experienced during the operation of PFR, PHENIX and KNK II, where 44 failed fuel elements have been registered between 1978 and 1989. In earlier irradiations, post irradiation examinations showed mixed oxide pin diameter increases up to pin pitch distance, urging to stress reactor safety questions on the potential of fuel pin failure propagation within pin bundles. The chemical interaction of sodium with mixed oxide fuel is regarded to be the key for the understanding of failed fuel behavior. Valuable results on the failed fuel pin behavior during operation were obtained from the SILOE sodium loop test. Based on the bulk of experience with the detection of fuel pin failures, with the continued operation and with the handling of failed pins respectively elements, one can state: 1. All fuel pin failures have been detected securely in time and have been located. 2. Small defects are developing slowly. 3. Even large defects at end-of-life pins resulted in limited fuel loss. 4. Clad failures behave benign in main aspects. 5. The chemical interaction of sodium with mixed oxide is an important factor in the behavior of failed fuel pins, especially at high burnup. 6. Despite different pin designs and different operation conditions, on the basis of 44 failed elements in PFR, PHENIX and KNK II no pin-to-pin propagation was observed and fuel release was rather low, often not detectable. 7. In no case hazard conditions affecting reactor safety have been experienced

Detection device for the failed position in fuels

International Nuclear Information System (INIS)

Tokunaga, Kensuke; Nomura, Teiji; Hiruta, Koji

1985-01-01

Purpose: To detect the failed position of a fuel assembly with ease and safety. Constitution: A fuel assembly is tightly closed in a sipper tube equipped with a gas supply tube and a gas exhaust tube at the upper portion and a purified water injection tube and a draining tube at the lower end. Then, water in the sipper tube is drained to the lower portion of the fuel assembly by the pressure of gases while opening the gas supply tube and the draining tube, and closing the exhaust tube and the injection tube. Then, after closing the gas supply tube and the draining tube while opening theexhaust tube and the injection tube, purified water is injected into the sipper tube from the injection tube to an optional height till the fuel assembly is immersed. Then, after leaving for a predetermined of time, water is sampled and the radioactive material density therein is measured. By changing the injection level of the purified water, since the radioactive material density changes at the failed position, the failed position can be detected with ease. (Sekiya, K.)

Post-irradiation examination of a failed PHWR fuel bundle of KAPS-2

International Nuclear Information System (INIS)

Mishra, Prerna; Unnikrishnan, K.; Viswanathan, U.K.; Shriwastaw, R.S.; Singh, J.L.; Ouseph, P.M.; Alur, V.D.; Singh, H.N.; Anantharaman, S.; Sah, D.N.

2006-08-01

Detailed post irradiation examination was carried out on a PHWR fuel bundle irradiated at Kakrapar Atomic Power Station unit 2 (KAPS-2). The fuel bundle had failed early in life at a low burnup of 387 MWd/T. Non destructive and destructive examination was carried out to identify the cause of fuel failure. Visual examination and leak testing indicated failure in two fuel pins of the outer ring of the bundle in the form of axial cracks near the end plug location. Ultrasonic testing of the end cap weld indicated presence of lack of fusion type defect in the two fuel pins. No defect was found in other fuel pins of the bundle. Metallographic examination of fuel sections taken from the crack location in the failed fuel pin showed extensive restructuring of fuel. The centre temperature of the fuel had exceeded 1700 degC at this location in the failed fuel pin, whereas fuel centre temperature in the un-failed fuel pin was only about 1300 degC. Severe fuel clad interaction was observed in the failed fuel pin at and near the location of failure but no such interaction was observed in the un-failed fuel pins. Several incipient cracks originating from the inside surface were found in the cladding near failure location in addition to the main through wall crack. The incipient cracks were filled with interaction products and hydride platelets were present at tip of the cracks. It was concluded from the observations that the primary cause of failure was the presence of a part-wall defect in the end cap weld of the fuel pins. These defects opened up during reactor operation leading to steam ingress into the fuel, which caused high fuel centre temperature and severe fuel-cladding interaction resulting in secondary failures. A more stringent inspection and quality control of end plug weld during fabrication using ultrasonic test has been recommended to avoid such failure. (author)

Failed fuel detection and location of LMFBR

International Nuclear Information System (INIS)

Mimoto, Yasuhide; Hukuda, Tooru; Nakamoto, Koichiro

1974-01-01

This is a summary report on Failed Fuel Detection and Location Methods of liquid metal cooled fast breeder reactors, and describes an outline of related research and development conducted by PNC. (auth.)

Detection method of a failed fuel

International Nuclear Information System (INIS)

Urata, Megumu; Uchida, Shunsuke; Utamura, Motoaki.

1976-01-01

Object: To divide a tank arrangement into a heating tank for the exclusive use of heating and a mixing tank for the exclusive use of mixing to thereby minimize the purifying amount of reactor water pumped from the interior of reactor and to considerably minimize the capacity of a purifier. Structure: In a detection method of a failed fuel comprising stopping a flow of coolant within fuel assemblies arranged in the coolant in a reactor container, sampling said coolant within the fuel assemblies, and detecting a radioactivity level of sampling liquid, the improvement of the method comprising the steps of heating a part of said coolant removed from the interior of said reactor container, mixing said heated coolant into the remainder of said removed coolant, pouring said mixed liquid into said fuel assemblies, and after a lapse of given time, sampling the liquid poured into said fuel assemblies. (Kawakami, Y.)

Detector for failed fuel elements

International Nuclear Information System (INIS)

Ito, Masaru.

1979-01-01

Purpose: To provide automatic monitor for the separation or reactor water and sampling water, in a failed fuel element detector using a sipping chamber. Constitution: A positional detector for the exact mounting of a sipping chamber on a channel box and a level detector for the detection of complete discharge of cooling water in the sipping chamber are provided in the sipping chamber. The positional detector is contacted to the upper end of the channel box and operated when the sipping chamber is correctly mounted to the fuel assemblies. The level detector comprises a float and a limit switch and it is operated when the water in the sipping chamber is discharged by a predetermined amount. Isolation of reactor water and sampling water are automatically monitored by the signal from these two detectors. (Ikeda, J.)

Radiolysis effects on fuel corrosion within a failed nuclear waste container

International Nuclear Information System (INIS)

Sunder, S.; Shoeshmith, D.W.; Christensen, H.C.

2003-01-01

The concept of geological disposal of used nuclear fuel in corrosion resistant containers is being investigated in several countries. In the Canadian Nuclear Fuel Waste Management Program (CNFWMP), it is assumed that the used fuel will be disposed of in copper containers. Since the predicted lifetimes of these containers are very long (>106 years), only those containers emplaced with an undetected defect will fail within the period for which radionuclide release from the fuel must be considered. Early failure could lead to the entry of water into the container and subsequent release of radionuclides. The release rate of radionuclides from the used fuel will depend upon its dissolution rate. The primary mechanism for release will be the corrosion of the fuel driven by radiolytically-produced oxidants. The studies carried out to determine the effects of water radiolysis on fuel corrosion are reviewed, and some of the procedures used to predict corrosion rates of used fuel in failed nuclear waste containers described. (author)

Transportation of failed or damaged foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

1999-01-01

Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

The cost of operating with failed fuel at Virginia power

International Nuclear Information System (INIS)

Ford, C.A.

1988-01-01

Virginia Power has completed a study of the costs incurred due to fuel failures in its pressurized water reactors. This study was prompted by histories of high primary coolant activity and subsequent fuel inspections at the North Anna and Surry power stations. The study included an evaluation of the total costs of fuel failures as well as an evaluation of the economics of postirradiation fuel inspections. The major costs of fuel failures included personnel radiation exposure, permanently discharged failed fuel, radwaste generation, increased labor requirements, containment entry delays due to airborne radioactivity, and ramp rate restrictions. Although fuel failures affect a utility in several other areas, the items evaluated in the study were thought to be the most significant of the costs. The study indicated that performing a postirradiation failed fuel examination can be economically justified at tramp-corrected 131 I levels of > 0.015 μCi/g. The savings to the utility can be on the order of several million dollars. Additionally, the cost penalty of performing a fuel inspection at lower iodine levels is generally in the range of $200,000. This economic penalty is expected to be outweighed by the intangible benefits of operating with a defect-free core

Transportation of failed or damaged foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, C.E.; Mustin, T.P.; Massey, C.D.

1998-01-01

Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

Sipping test on a failed MTR fuel element

International Nuclear Information System (INIS)

Terremoto, Luis Antonio Albiac; Zeituni, Carlos Alberto; Silva, Antonio Teixeira e; Perrotta, Jose Augusto; Silva, Jose Eduardo Rosa da

2002-01-01

This work describes sipping tests performed on MTR fuel elements of the IEA-R1 research reactor, in order to determinate which one failed in the core during a routine operation of the reactor. radioactive iodine isotopes 131 I and 133 I, employed as failure indicators, were detected in samples corresponding to the fuel element IEA-156. The specific activity of each sample, as well as the average leaking rate, were measured for 137 Cs. The nuclear fuels U 3 O 8 - Al dispersion and U - Al alloy were compared concerning their measured average leaking rates of 137 Cs. (author)

Integrated identification and control for nanosatellites reclaiming failed satellite

Science.gov (United States)

Han, Nan; Luo, Jianjun; Ma, Weihua; Yuan, Jianping

2018-05-01

Using nanosatellites to reclaim a failed satellite needs nanosatellites to attach to its surface to take over its attitude control function. This is challenging, since parameters including the inertia matrix of the combined spacecraft and the relative attitude information of attached nanosatellites with respect to the given body-fixed frame of the failed satellite are all unknown after the attachment. Besides, if the total control capacity needs to be increased during the reclaiming process by new nanosatellites, real-time parameters updating will be necessary. For these reasons, an integrated identification and control method is proposed in this paper, which enables the real-time parameters identification and attitude takeover control to be conducted concurrently. Identification of the inertia matrix of the combined spacecraft and the relative attitude information of attached nanosatellites are both considered. To guarantee sufficient excitation for the identification of the inertia matrix, a modified identification equation is established by filtering out sample points leading to ill-conditioned identification, and the identification performance of the inertia matrix is improved. Based on the real-time estimated inertia matrix, an attitude takeover controller is designed, the stability of the controller is analysed using Lyapunov method. The commanded control torques are allocated to each nanosatellite while the control saturation constraint being satisfied using the Quadratic Programming (QP) method. Numerical simulations are carried out to demonstrate the feasibility and effectiveness of the proposed integrated identification and control method.

Method and device for detecting failed fuels

International Nuclear Information System (INIS)

Saito, Shozo; Suzumura, Takeshi.

1981-01-01

Purpose: To shorten the time required for inspecting a failed fuel by providing a first outlet for exhausting cleaning liquid to a sampling pipe and a second outlet for exhausting sampled coolant, thereby safely setting a collecting means to the first outlet. Constitution: A sampling pipe is inserted into a fuel assembly loaded within a reactor core, and coolant flow is thus prevented from passing through the fuel assembly interior. Then, with the coolant flow stopped, it is allowed to stand for a predetermined time. Subsequently, cleaning liquid is supplied into the sampling tube and the interior of the sampling pipe is cleaned. Thereafter, the sampling liquid that was in the sampling pipe is exhausted from the first outlet of the sampling pipe. Then, the coolant in the fuel assembly is supplied from the second outlet of the sampling pipe to a collecting means. (Aizawa, K.)

Field experience with failed-fuel detection - PWRs [pressurized water reactors

International Nuclear Information System (INIS)

File, P.

1990-01-01

Industry sensitivity to in-reactor fuel reliability has been heightened recently for several reasons: increased Institute of Nuclear Power Operations attention (fuel reliability indicator), impact of fuel failures on operating and maintenance costs, and concern about personnel contaminations caused by small radioactive particles (often fragments from failed rods). As a result, utilities often perform inspection and repair campaigns at the end of a cycle where fuel failures have been evident in order to avoid experiencing the effects of those failures during the subsequent cycle. At Calvert Cliffs, the first full-core inspection associated with concern over fuel integrity occurred in early 1984 after unit 2 cycle 5. Lessons learned from this campaign are discussed in this paper. While significant improvement has been made relative to the ability to discern failed rods from sound rods, ultrasonic testing should be evaluated by people experienced with mechanisms that can produce questionable indications. The speed and accuracy of UT systems make UT a practical tool for performing inspections with minimal impact on the schedule

Fuel number identification method and device

International Nuclear Information System (INIS)

Doi, Takami; Seno, Makoto; Kikuchi, Takashi; Sakamoto, Hiromi; Takahashi, Masaki; Tanaka, Keiji.

1997-01-01

The present invention provides a method of and a device for automatically identifying fuel numbers impressed on fuel assemblies disposed in a fuel reprocessing facility, power plant and a reactor core at a high speed and at a high identification rate. Namely, three or more character images are photographed for one fuel assembly as an object of the identification under different illumination conditions. As a result, different character images by the number of the illumination directions can be obtained for identical impressed characters. Learning on a neural network system is applied to the images of all of the characters impressed on the fuel assembly obtained under illumination of predetermined directions. Then, result of the identification by the number of the illumination directions can be obtained for each of the characters as an object of the identification. As a result, since the result of the identification is determined based on a theory of decision of majority, highly automatic identification can be realized. (I.S.)

Aviation Fuel System Reliability and Fail-Safety Analysis. Promising Alternative Ways for Improving the Fuel System Reliability

Directory of Open Access Journals (Sweden)

I. S. Shumilov

2017-01-01

Full Text Available The paper deals with design requirements for an aviation fuel system (AFS, AFS basic design requirements, reliability, and design precautions to avoid AFS failure. Compares the reliability and fail-safety of AFS and aircraft hydraulic system (AHS, considers the promising alternative ways to raise reliability of fuel systems, as well as elaborates recommendations to improve reliability of the pipeline system components and pipeline systems, in general, based on the selection of design solutions.It is extremely advisable to design the AFS and AHS in accordance with Aviation Regulations АП25 and Accident Prevention Guidelines, ICAO (International Civil Aviation Association, which will reduce risk of emergency situations, and in some cases even avoid heavy disasters.ATS and AHS designs should be based on the uniform principles to ensure the highest reliability and safety. However, currently, this principle is not enough kept, and AFS looses in reliability and fail-safety as compared with AHS. When there are the examined failures (single and their combinations the guidelines to ensure the AFS efficiency should be the same as those of norm-adopted in the Regulations АП25 for AHS. This will significantly increase reliability and fail-safety of the fuel systems and aircraft flights, in general, despite a slight increase in AFS mass.The proposed improvements through the use of components redundancy of the fuel system will greatly raise reliability of the fuel system of a passenger aircraft, which will, without serious consequences for the flight, withstand up to 2 failures, its reliability and fail-safety design will be similar to those of the AHS, however, above improvement measures will lead to a slightly increasing total mass of the fuel system.It is advisable to set a second pump on the engine in parallel with the first one. It will run in case the first one fails for some reasons. The second pump, like the first pump, can be driven from the

Design study and comparative evaluation of JSFR failed fuel detection system

Energy Technology Data Exchange (ETDEWEB)

Aizawa, K.; Chikazawa, Y.; Ishikawa, N. [Japan Atomic Energy Agency JAEA, 4002 Narita, Oarai, Higashi-ibaraki-gun, Ibaraki 311-1393 (Japan); Kubo, S. [Japan Atomic Power Company JAPC (Japan); Okazaki, H.; Mito, M. [Mitsubishi FBR Systems, Inc. MFBR (Japan); Tozawa, K. [Fuji Electric Co., Ltd. (Japan); Hayashi, M. [MitsubishiElectric Corporation (Japan)

2012-07-01

A conceptual design study of an advanced sodium-cooled fast reactor JSFR has progressed in the 'Fast Reactor Cycle Technology Development (FaCT) 'project in Japan. JSFR has two failed fuel detection systems in the core. One is a failed fuel detection (FFD) system which continuously monitors a fission product from failed fuel subassembly. The other is a failed fuel detection and location (FFDL) system which locates when it receives signals from FFD. The FFD system consists of a FFD-DN which detects delayed neutron (DN) in sodium and a FFD-CG which detects fission products in the cover gas of the reactor vessel. In this study, requirements to the FFD-DN and the FFD-DN design to meet the requirements were investigated for the commercial and demonstration JSFR. In the commercial JSFR, a sampling type FFD which collects sodium from the reactor vessel by sampling lines for DN detectors was adopted. The performances have been investigated and confirmed by a fluid analysis in the reactor upper plenum. In the demonstration JSFR, the performance of DN detectors installed on the primary cold-leg piping has been confirmed. For the FFDL systems, experiences in the previous fast reactors and the R and D of FFDL system for JSFR were investigated. This study focuses on the Selector-Valve and the Tagging-Gas FFDL systems. Operation experiences of the Selector-valve FFDL system were accumulated in PFR and Phenix. Tagging-gas system experiences were accumulated in EBR-II and FFTF. The feasibility of both FFDL systems for JSFR was evaluated. (authors)

Categorization of failed and damaged spent LWR [light-water reactor] fuel currently in storage

International Nuclear Information System (INIS)

Bailey, W.J.

1987-11-01

The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs

Electro-optical fuel pin identification system

International Nuclear Information System (INIS)

Kirchner, T.L.

1978-09-01

A prototype Electro-Optical Fuel Pin Identification System referred to as the Fuel Pin Identification System (FPIS) has been developed by the Hanford Engineering Development Laboratory (HEDL) in support of the Fast Flux Test Facility (FFTF) presently under construction at HEDL. The system is designed to remotely read an alpha-numeric identification number that is roll stamped on the top of the fuel pin end cap. The prototype FPIS consists of four major subassemblies: optical read head, digital compression electronics, video display, and line printer

Monte Carlo modeling for realizing optimized management of failed fuel replacement

International Nuclear Information System (INIS)

Morishita, Kazunori; Yamamoto, Yasunori; Nakasuji, Toshiki

2014-01-01

Fuel cladding is one of the key components in a fission reactor to keep confining radioactive materials inside a fuel tube. During reactor operation, the cladding is however sometimes breached and radioactive materials leak from the fuel ceramic pellet into the coolant water through the breach. The primary coolant water is therefore monitored so that any leak is quickly detected, where the coolant water is periodically sampled and the concentration of, for example the radioactive iodine 131 (I-131), is measured. Depending on the measured concentration, the faulty fuel assembly with leaking rod is removed from the reactor and replaced by new one immediately or at the next refueling. In the present study, an effort has been made to develop a methodology to optimize the management for replacement of failed fuels due to cladding failures using the I-131 concentration measured in the sampled coolant water. A model numerical equation is proposed to describe the time evolution of I-131 concentration due to fuel leaks, and is then solved using the Monte-Carlo method as a function of sampling rate. Our results have indicated that, in order to achieve the rationalized management of failed fuels, higher resolution to detect a small amount of I-131 is not necessarily required but more frequent sampling is favorable. (author)

Characteristics of axial splits in failed BWR fuel rods

International Nuclear Information System (INIS)

Lysell, G.; Grigoriev, V.

2000-01-01

Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

A combined wet/dry sipping cell for investigating failed triga fuel elements

International Nuclear Information System (INIS)

Boeck, H.; Gallhammer, H.; Hammer, J.; Israr, M.

1987-08-01

A sipping cell to detect failed triga fuel has been designed and constructed at the Atominstitut. The cell allows both wet- and dry sipping of one single standard triga fuel element. In the dry sipping method the fuel element may be electrically heated up to a maximum temperature of about 300 0 C to allow the detection of temperature dependent fission product release from the fuel element. 20 figs., 1 tab. (Author)

Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

2006-01-01

The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

Acceptance of failed SNF [spent nuclear fuel] assemblies by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. This document discusses acceptance of failed spent fuel assemblies by the Federal Waste Management System. 18 refs., 7 figs., 25 tabs

77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

Science.gov (United States)

2012-03-05

... Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Withdrawal... Renewable Fuel Standard program regulations. Because EPA received adverse comment, we are withdrawing the...

Hydraulic experiments on the failed fuel location module of prototype fast breeder reactor

International Nuclear Information System (INIS)

Rajesh, K.; Kumar, S.; Padmakumar, G.; Prakash, V.; Vijayashree, R.; Rajan Babu, V.; Govinda Rajan, S.; Vaidyanathan, G.; Prabhaker, R.

2003-01-01

The design of Prototype Fast Breeder Reactor (PFBR) is based on sound design concepts with emphasis on intrinsic safety. The uncertainties involved in the design of various components, which are difficult to assess theoretically, are experimentally verified before design is validated. In PFBR core, the coolant (liquid sodium) enters the bottom of the fuel subassembly, passes over the fuel pins picking up the fission heat and issues in to a hot pool. If there is any breach in the fuel pins, the fission products come in direct contact with the coolant. This is undesirable and it is necessary to locate the subassembly with the failed fuel pin and to isolate it. A component called Failed Fuel Location Module (FFLM) is employed for locating the failed SA by monitoring the coolant samples coming out of each Subassembly. The coolant sample from each Subassembly is drawn by FFLM using an EM pump through sampling tube and selector valve and is monitored for the presence of delayed neutrons which is an indication of failure of the Subassembly. The pressure drop across the selector valve determines the rating of the EM Pump. The dilution of the coolant sample across the selector valve determines the effectiveness of monitoring for contamination. It is not possible to predict pressure drop across the selector valve and dilution of the coolant sample theoretically. These two parameters are determined using a hydraulic experiment on the FFLM. The experiment was carried out in conditions that simulate the reactor conditions following appropriate similarity laws. The paper discusses the details of the model, techniques of experiments and the results from the studies

Method and apparatus for detecting failed fuels in nuclear reactor

International Nuclear Information System (INIS)

Tsuji, Tadashi.

1981-01-01

Purpose: To enable safety and automatic sampling for sample water in failed fuel detection. Constitution: A cap containing inner caps by the number of fuel assemblies inserted into each grid of a nuclear reactor is mounted to the upper end of the fuel assemblies. After the mounting, it is confirmed if the mounting is collectly made by the mounting state detection device utilizing the change in the pressure within the tube communicating to a water seal pipe. Then, air at a predetermined pressure introduced from an air supply tube opening into the cap is introduced into the cap to replace the coolants in the cap with the air. The pressure difference between the inside and the outside of the cap is detected and, if it shows a set value, it is confirmed that the cooling water level is independent for every fuel assembly. Then, sample water is sampled by the sampling tube within the guide cap provided to the upper end of the inner cap. (Furukawa, Y.)

Neural signal processing for identifying failed fuel rods in nuclear reactors

International Nuclear Information System (INIS)

Seixas, Jose M. de; Soares Filho, William; Pereira, Wagner C.A.; Teles, Claudio C.B.

2002-01-01

Ultrasonic pulses were used for automatic detection of failed nuclear fuel rods. For experimental tests of the proposed method, an assembly prototype of 16 x 16 rods was built by using genuine rods but without fuel inside (just air). Some rods were partially filled with water to simulate cracked rods. Using neural signal processing on the received echoes of the emitted ultrasonic pulses, a detection efficiency of 97% was obtained. Neural detection is shown to outperform other classical discriminating methods and can also reveal important features of the signal structure of the received echoes. (author)

Failed (leaking) spent fuel management and storage in the Paks NPP

International Nuclear Information System (INIS)

Burjan, T.

2011-01-01

At the cycle 22, unit 4, Paks NPP the fissile contents raised irregularly in the water of the primary circuit. At the end of the cycle sipping tests were performed for the entire core to find out the leaking fuel assembly primarily responsible for this phenomenon. The identified leaking assembly temporarily was placed in the Spent Fuel Relaxing Pool. For measuring environmental impact of leaking assemblies an investigation program was developed and implemented. The assessment covered the following: effects of the leaking fuel on the water of relaxing pool and on the gaseous emissions in case open storage; in case when the leaking cassette is in a special hermetical storage case, how much gas is collected in the locked case and what is its composition; how to change the measured sipping test signal depending on relaxing time of leaking fuel cassettes. Based on the evaluation of the investigation program results the NPP modified the operational instructions for the treatment and storage of failed fuel assemblies. (author)

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Fuel number identification method and device therefor

International Nuclear Information System (INIS)

Doi, Takami; Seno, Makoto; Tanaka, Keiji

1998-01-01

The present invention provides a method of and a device for automatically identifying the number on the upper surface of a fuel of a fuel assembly in a PWR type reactor. Namely, the number on the upper surface of the fuel assembly of the PWR is not arranged in a row, but indent letters are dispersed to predetermined positions of the surface to be indented. Accordingly, the identification of letters is difficult. In the present invention, the letters are identified by the following procedures. Procedure (1): the letters are detected while having a corner portion of the upper surface of a fuel assembly where the number is indented as characteristic points. A procedure (2): a letter region is determined to a relative position based on the characteristic points while determining indent letters having the same direction as one group. A procedure (3): a letter identification treatment is applied to the letter images in the above-mentioned letter region to identify them. A neural network is used for the letter identification treatment. (N.H.)

Criteria for recladding of spent light water reactor fuel before long term pool storage

International Nuclear Information System (INIS)

Pettersson, K.; Jansson, L.

1979-01-01

The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that these degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods. (Auth.)

Criteria for recladding of spent light water reactor fuel before long term pool storage

International Nuclear Information System (INIS)

Pettersson, K.; Jansson, L.

1979-06-01

The question of the need for any special treatment of failed fuel elements prior to long term pool storage has been studied. It is concluded that the main problem appears to be hydride embrittlement of failed fuel rods, which may lead to increased damage during handling and transport of the failed fuel. Some mechanisms for the degradation of failed fuel rods have been identified. They can all be considered as relatively improbable, but further experimental evidence is needed before it can be concluded that thede degradation mechanisms are insignificant during pool storage. The report also contains a review of methods for identification of leaking fuel bundles and fuel rods.(author)

Failed fuel rod detection system and computerized manipulator during outages

International Nuclear Information System (INIS)

Boehm, H.H.; Foerch, H.

1984-01-01

During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

Convective-diffusive transport of fission products in the gap of a failed fuel element

International Nuclear Information System (INIS)

Lian, Z.W.; Carlucci, L.N.; Arimescu, V.I.

1995-03-01

A model is presented to describe the transport behaviour of gaseous fission products along the axial fuel-to-sheathe gap of a failed fuel element to the coolant system. The model is applicable to an element having failed under normal operating conditions or loss-of coolant-accident conditions. Because of the large differences in operating parameters, the transport characteristics of gaseous fission products in a failed element under these two operating conditions are significantly different. However, in both cases the transport process can be described by convection-diffusion caused by the continuous release of fission products from the fuel to the gap. Under normal operating conditions, the bulk-flow velocity is found to be negligible, due to the low release rate of fission products from fuel. The process can be well approximated by the diffusion of fission products in a stagnant gas-steam mixture. The effect of convection on the fission product transport, however, becomes significant under loss-of-coolant-accident conditions, where the release rates of fission products from fuel can be several orders of magnitude higher that that under normal operating conditions. The convection of the mixture in the gap not only contributes an additional flux to the gas-mixture transport, but also increases the gradient of fission products concentration across the opening, and therefore increases the diffusion flux to the coolant. As a result of the bulk flow, the transport of fission products along the gap is accelerated and the hold-up of short-lived isotopes in the gap is significantly reduced. Steam ingress through the opening into the gap is obstructed by the bulk flow, resulting in low steam concentrations in the gap under loss-of-coolant-accident conditions. (author). 6 refs., 8 figs

A follow-up test of failed fuel element of a nuclear reactor

International Nuclear Information System (INIS)

Peerasathien, W.

1974-01-01

This thesis is a result of test of a number of nuclear fuel rods which have not been used for a long time due to leakage of radioactivity. Water was circulated through each fuel rod in a test cylinder and radioactivity in water was measured. It was found that the detection of Cesium-137 which has a long half-life, does not indicate the extent of leakage of short-lived radioisotopes, some of which are gaseous. These gases are harmful to the reactor operators and users. A better result was obtained by placing the failed fuel rod in the test cylinder close to the reactor to induce fission. Short half-life gases or other nuclides of the same series were then directly measured

Pressurised light water reactor failed fuel monitoring. Comparison of international strategies used in the surveillance of the primary barrier

International Nuclear Information System (INIS)

Aldworth, Robin; Dinse, Christelle; Mole, Didier

2014-01-01

As part of its focus on continuous improvement, the nuclear industry has devoted considerable effort to the prevention, monitoring, characterisation and impact mitigation of fuel failures. Failed fuel and primary circuit activity monitoring strategies have thus evolved over time, with utilities changing their strategies as a function of operational experience, regulator encouragement and improvements to scientific understanding. Differences among utility fuel monitoring strategies are therefore to be expected. This paper compares and analyses a selection of failed fuel and primary circuit activity monitoring strategies employed at French, American, South-African and Belgian pressurised water reactors (PWRs). These strategies, which were received in response to a questionnaire sent by Electricité de France, are discussed with a number of potential best practices and areas for future research and development being provided. Finally, and so as to ensure the utility of this paper with regards to the development, or evolution of, failed fuel monitoring strategies, this paper has been split into four sections representing the primary radiochemistry indicator groups: gross, calculated, fission products and activated corrosion products. (author)

Studies on modeling to failed fuel detection system response in LMFBR

International Nuclear Information System (INIS)

Miyazawa, T.; Saji, G.; Mitsuzuku, N.; Hikichi, T.; Odo, T.; Rindo, H.

1981-05-01

Failed Fuel Detection (FFD) system with Fission Products (FP) detection is considered to be the most promissing method, since FP provides direct information against fuel element failure. For designing FFD system and for evaluating FFD signals, some adequate FFD signal response to fuel failure have been required. But few models are available in nowadays. Thus Power Reactor and Nuclear Fuel Development Corporation (PNC) had developed FFD response model with computer codes, based on several fundamental investigations on FP release and FP behavior, and referred to foreign country experiences on fuel failure. In developing the model, noble gas and halogen FP release and behavior were considered, since FFD system would be composed of both cover gas monitoring and delayed neutron monitoring. The developed model can provide typical fuel failure response and detection limit which depends on various background signals at cover gas monitoring and delayed neutron monitoring. According to the FFD response model, we tried to assume fuel failure response and detection limit at Japan experimental fast reactor ''JOYO''. The detection limit of JOYO FFD system was estimated by measuring the background signals. Followed on the studies, a complete computer code has been now made with some improvement. On the paper, the details of the model, out line of developed computer code, status of JOYO FFD system, and trial assumption of JOYO FFD response and detection limit. (author)

Used Fuel Cask Identification through Neutron Profile

Energy Technology Data Exchange (ETDEWEB)

Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-11-20

Currently, most spent fuel is stored near reactors. An interim consolidated fuel storage facility would receive fuel from multiple sites and store it in casks on site for decades. For successful operation of such a facility there is need for a way to restore continuity of knowledge if lost as well as a method that will indicate state of fuel inside the cask. Used nuclear fuel is identifiable by its radiation emission, both gamma and neutron. Neutron emission from fission products, multiplication from remaining fissile material, and the unique distribution of both in each cask produce a unique neutron signature. If two signatures taken at different times do not match, either changes within the fuel content or misidentification of a cask occurred. It was found that identification of cask loadings works well through the profile of emitted neutrons in simulated real casks. Even casks with similar overall neutron emission or average counts around the circumference can be distinguished from each other by analyzing the profile. In conclusion, (1) identification of unaltered casks through neutron signature profile is viable; (2) collecting the profile provides insight to the condition and intactness of the fuel stored inside the cask; and (3) the signature profile is stable over time.

Integration of post-irradiation examination results of failed WWER fuel rods

International Nuclear Information System (INIS)

Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Perepelkin, S.

2003-01-01

The aim of the work is to investigate the causes of WWER fuel rod failures and to reveal the dependence of the failed fuel rod behaviour and state on the damage characteristics and duration of their operation in the core. The post-irradiation examination of 12 leaky fuel assemblies (5 for WWER-440 and 7 for WWER-1000) has been done at SSC RF RIAR. The results show that the main mechanism responsible for the majority of cases of the WWER fuel rod perforation is debris-damage of the claddings. Debris fretting of the claddings spread randomly over the fuel assembly cross-section and they are registered in the area of the bundle supporting grid or under the lower spacer grids along the fuel assembly height. In the WWER fuel rods, the areas of secondary hydrogenating of cladding are spaced from the primary defects by ∼2500-3000 mm, as a rule, and are often adjacent closely to the upper welded joints. There is no pronounced dependence of the distance between the primary and secondary cladding defects neither on the linear power, at which the fuel rods were operated, nor on the period of their operation in the leaky state. The time period of the significant secondary damage formation is about 250 ± 50 calendar days for the WWER fuel rods with slight through primary defects (∼0.1 - 0.5 mm 2 ) operated in the linear power range 170-215 W/cm. Cladding degradation, taking place due to the secondary hydrogenating, does not occur in case of large through debris-defects during operation up to 600 calendar days

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

OpenAIRE

Asnavandi, Majid; Kahram, Mohaddeseh; Rezaei, Milad; Rezakhani, Davar

2017-01-01

The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tu...

Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

International Nuclear Information System (INIS)

Kryter, R.C.

1975-01-01

Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

Comparison of biochemical and molecular methods for the identification of bacterial isolates associated with failed loggerhead sea turtle eggs.

Science.gov (United States)

Awong-Taylor, J; Craven, K S; Griffiths, L; Bass, C; Muscarella, M

2008-05-01

Comparison of biochemical vs molecular methods for identification of microbial populations associated with failed loggerhead turtle eggs. Two biochemical (API and Microgen) and one molecular methods (16s rRNA analysis) were compared in the areas of cost, identification, corroboration of data with other methods, ease of use, resources and software. The molecular method was costly and identified only 66% of the isolates tested compared with 74% for API. A 74% discrepancy in identifications occurred between API and 16s rRNA analysis. The two biochemical methods were comparable in cost, but Microgen was easier to use and yielded the lowest discrepancy among identifications (29%) when compared with both API 20 enteric (API 20E) and API 20 nonenteric (API 20NE) combined. A comparison of API 20E and API 20NE indicated an 83% discrepancy between the two methods. The Microgen identification system appears to be better suited than API or 16s rRNA analysis for identification of environmental isolates associated with failed loggerhead eggs. Most identification methods are not intended for use with environmental isolates. A comparison of identification systems would provide better options for identifying environmental bacteria for ecological studies.

Special storage of leaking fuel at Paks NPP

International Nuclear Information System (INIS)

Biro, Janos; Szőke, L.; Burján, T.; Lukács, R.; Hózer, H.

2015-01-01

In this paper the activities related with spent, hermetic as well as leaking fuel handling and storage, including: Spent fuel pool; Transportation criteria for the spent fuel assemblies and Interim spent fuel dry storage; Short-term storage in the spent fuel pool; Identification of the leaking assemblies by the TS-device; Present conception of Identification, handling of the leaking FAs; Modified transport procedure for the leaking FAs; Calculation of solved activity inside the leaking fuel rod; Solved activity limit values for the leaking FAs; Long-term storage in the interim spent Fuel dry storage are presented. At the end authors’ concluded that: 1) The leaking FA can be transported to the interim dry storage together with the other spent fuel assemblies in the transport container. 2) The transport-documentation of the leaking FA has to contain: isotope inventory, calculated solved activity values of the failed FA and the quantity of failed fuel rods. 3) Performing three leakage tests of the identified leaking FA before the transportation in the 5FP. it is useful to decrease the solved activity concentration inside the leaking FA and give additional information about the extent of the leakage. 4) We can calculate simply the solved activity of the leaking FA. 5) The modified transport procedure will have to be authorized. 6) The radiological effects of the leaking FA are negligible relative to the natural background radiation

Fail-safe system for activity cooled supersonic and hypersonic aircraft. [using liquid hydrogen fuel

Science.gov (United States)

Jones, R. A.; Braswell, D. O.; Richie, C. B.

1975-01-01

A fail-safe-system concept was studied as an alternative to a redundant active cooling system for supersonic and hypersonic aircraft which use the heat sink of liquid-hydrogen fuel for cooling the aircraft structure. This concept consists of an abort maneuver by the aircraft and a passive thermal protection system (TPS) for the aircraft skin. The abort manuever provides a low-heat-load descent from normal cruise speed to a lower speed at which cooling is unnecessary, and the passive TPS allows the aircraft skin to absorb the abort heat load without exceeding critical skin temperature. On the basis of results obtained, it appears that this fail-safe-system concept warrants further consideration, inasmuch as a fail-safe system could possibly replace a redundant active cooling system with no increase in weight and would offer other potential advantages.

Aims of failed fuel detection and substantiation of radiation safety at implementation of new kinds of nuclear fuel and fuel cycles on NPP with WWER

International Nuclear Information System (INIS)

Miglo, V.; Luzanova, L.

2011-01-01

Limiting of number of leaking fuel rods in a core during reactor operation in the analyses which are carried out for a substantiation of radiating safety for NPP with WWER as well as problems and possibilities of FFD at implementation of new kinds of fuel and fuel cycles are the main topics discussed in this paper. Available experience of designing of the NPP with WWER shows, that for ensuring of implementation of the RS criteria regarding limiting radioactive emissions from the NPP and doses of an irradiation of the population living near to NPP, it is required to regulate more rigidly number of failed fuel rods in comparison with requirements of Rules of nuclear safety NP-082-07. The reason of it is necessity to consider a technical condition of all safety barriers on a path of radioactive FP extension in a complex, first and foremost of uncontrolled leakage of the primary coolant to the NPP premises and efficiency of filters of ventilating systems, and also spike-effect on activity of isotopes of iodine after a power unit shutdown for fuel reloading and openings of a cover of a reactor. Depending on the project of NPP, parameters of fuel loading, a place of placing of the NPP and other factors the limit level of activity of isotopes of iodine in the primary coolant will be reached at various number of leaking fuel rods which can be unequal for various power units and the NPP with WWER, constructed on one design. The quantity of leaking fuel rods at which the design limit on FP-activity in the primary coolant of operating reactor is reached, can be essential below an operational limit on number of failed fuel rods established by Rules of nuclear safety. However the reached quality of fabrication of the WWER fuel rods providing their high reliability (the probability of fuel rod failure in the course of one operation year is not higher than 10 -5 ) as well as due to the levels of the WWER fuel rod depressurization actually attainable in the normal conditions of

Studying some regimes of the WWER-440 type reactor failed fuel element operation

International Nuclear Information System (INIS)

Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

1981-01-01

The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Energy Technology Data Exchange (ETDEWEB)

Hunn, John D., E-mail: hunnjd@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-09-15

Highlights: • Cesium release was used to detect SiC failure in HTGR fuel. • Tristructural-isotropic particles with SiC failure were isolated by gamma screening. • SiC failure was studied by X-ray tomography and SEM. • SiC degradation was observed after irradiation and subsequent safety testing. - Abstract: As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of {sup 134}Cs and {sup 137}Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were

Detection and Analysis of Particles with Failed SiC in AGR-1 Fuel Compacts

International Nuclear Information System (INIS)

Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

2014-01-01

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of "1"3"4Cs and "1"3"7Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during the AGR-1 irradiation test or post-irradiation safety testing at 1600– 1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the ORNL Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking. (author)

From targeting to implementation: The role of identification of fuel poor households

International Nuclear Information System (INIS)

Dubois, Ute

2012-01-01

This paper proposes to evaluate the efficiency of fuel poverty policies by looking at them as a three-step process. These steps are policy targeting, identification of households and implementation of measures. Each of these steps are costly and can face feasibility problems. This is particularly true in the case of fuel poverty policies, because of the multidimensional aspect of fuel poverty. Therefore, public policies in that field are necessarily imperfect, and this should be taken into account in the analysis of policies. This three-step framework is applied to the French fuel poverty policy, which has been developed in 2010. As its practical modalities of implementation are still under discussion, we mainly discuss the choices made in terms of targeting and highlight the role of identification of fuel poor households as a key factor for the success of that policy. The general idea behind this discussion is that the efficiency of such policies has to be viewed in a comprehensive way, which includes the three policy steps. - Highlights: ► This paper analyses fuel poverty policy as a three-step process. ► The steps are targeting, identification and implementation. ► Feasibility and implementation problems should be considered in policy formulation. ► Errors of inclusion and exclusion limit the effectiveness of fuel poverty policy. ► We apply this framework to the recent French fuel poverty policy.

Fundamental principles of failed fuel detection concepts on nuclear power units of WWER type

International Nuclear Information System (INIS)

Lusanova, L.; Miglo, V.; Slavyagin, P.

2001-01-01

The subject of the paper is the Russian failed fuel detection concept in both operating and shut down reactors. The philosophy for detection of fission products released from defective fuel during operation and sipping tests and using of these results for regulation of the radiological situation at the NPP during the next cycle is widely spread. In presented work such philosophy is applied to the shut down rectors. An option for sipping test performed in a mast of Refueling Machine (RM) using a wet-gas version of sipping test is briefly described. The use of the FFD method in RM mast allows combining the procedure of Fuel Assemblies (FA) tightness test with transport operation during reloading of the fuel from the core into the cooling pool. This reduces the time for reloading and transport operation with FA and increases the safety of reactor operation. The FFD method in RM mast has passed successful tests on Unit 4 at Balakovskaja NPP and it is expected to apply in other NPP unit with WWER-1000 reactors

Method of detecting failed fuels for a reactor

International Nuclear Information System (INIS)

Urata, Megumu; Kamiya, Kunio.

1982-01-01

Purpose: To detect the failed fuel with high sensitivity without causing insensitive measuring zone and measuring efficiency decrease time zone even if the temperature and the flow rate of sampling gas vary. Constitution: The simultaneous ventilation times t 2 of activated charcoal adsorption towers A and B are set to a value less than the holding-up time of primary radioactive nuclide in the one tower A, more than the time reaching the ordinary flowout state of secondary radioactive nuclide in the other tower B and yet less than 1/2 of the remaining time t 6 obtained by subtracting the switching time t 1 of the one tower A by the retreating time t 5 of the adsorbent. In the case that the ventilation to the one tower A and the measurement of the radioactivity level is switched to the other tower after the simultaneous ventilation in this manner, the flowout of the secondary radioactive nuclide in the other tower B becomes the ordinary state. Accordingly, no measuring efficiency decreasing time zone and insensitive measurement zone will not take place. (Seki, T.)

Failed fuel monitoring at nuclear power plants with RBMK reactors: operating parameters, requirements and decision making criteria

International Nuclear Information System (INIS)

Zhukov, I.V.

1993-01-01

The procedure for estimating the number of failed fuel rods in the core and the prediction of their discharge efficiency during operation is presented. The procedure is based on the FFM data base and the I-131 and Xe-133 coolant activity. (author)

Operational experience gained with the failed fuel rod detection system in nuclear power plants

International Nuclear Information System (INIS)

Boehm, H.H.; Forch, H.

1985-01-01

Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

Detection of failed fuel rods in shrouded BWR fuel assemblies

International Nuclear Information System (INIS)

Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

1988-01-01

A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

Contamination of a PWR primary circuit by fuel pins with failed cladding

International Nuclear Information System (INIS)

Janvier, J.C.; Chagrot, M.

1979-01-01

The safety authorities in the principal nuclear countries appear to be attaching increasing importance to keeping reactor primary circuits as contamination-free as possible. Therefore, the consequences of cladding failures and especially of those resulting from fabrication defects have to be evaluated, for when these failures become systematic in nature they constitute an important source of contamination in pressurized-water reactors. The Grenoble Nuclear Research Centre is implementing a programme on the study of such failures with a view to analysing the behaviour of failed fuel elements. A distinction is made between two types of cladding failure, depending on whether the primary water enters the fuel pin as soon as the circuits are pressurized (fabrication defect) or whether the failure is caused during operation. The emission of gaseous fission products and halogens has been analysed in different operating modes (steady-state or transient), and in spite of the complexity of the phenomena involved, some results have been obtained which already enable one to evaluate fission product contamination of the primary circuit. (author)

Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

Energy Technology Data Exchange (ETDEWEB)

Strindehag, O

1966-06-15

Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type ({nu} = 10{sup -8}/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the {gamma}-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A{sup 41} in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems.

Unaware person recognition from the body when face identification fails.

Science.gov (United States)

Rice, Allyson; Phillips, P Jonathon; Natu, Vaidehi; An, Xiaobo; O'Toole, Alice J

2013-11-01

How does one recognize a person when face identification fails? Here, we show that people rely on the body but are unaware of doing so. State-of-the-art face-recognition algorithms were used to select images of people with almost no useful identity information in the face. Recognition of the face alone in these cases was near chance level, but recognition of the person was accurate. Accuracy in identifying the person without the face was identical to that in identifying the whole person. Paradoxically, people reported relying heavily on facial features over noninternal face and body features in making their identity decisions. Eye movements indicated otherwise, with gaze duration and fixations shifting adaptively toward the body and away from the face when the body was a better indicator of identity than the face. This shift occurred with no cost to accuracy or response time. Human identity processing may be partially inaccessible to conscious awareness.

40 CFR 80.521 - What are the standards and identification requirements for diesel fuel additives?

Science.gov (United States)

2010-07-01

... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Motor... consumer in diesel motor vehicles or nonroad diesel engines. [69 FR 39171, June 29, 2004] ... identification requirements for diesel fuel additives? 80.521 Section 80.521 Protection of Environment...

Accelerating parameter identification of proton exchange membrane fuel cell model with ranking-based differential evolution

International Nuclear Information System (INIS)

Gong, Wenyin; Cai, Zhihua

2013-01-01

Parameter identification of PEM (proton exchange membrane) fuel cell model is a very active area of research. Generally, it can be treated as a numerical optimization problem with complex nonlinear and multi-variable features. DE (differential evolution), which has been successfully used in various fields, is a simple yet efficient evolutionary algorithm for global numerical optimization. In this paper, with the objective of accelerating the process of parameter identification of PEM fuel cell models and reducing the necessary computational efforts, we firstly present a generic and simple ranking-based mutation operator for the DE algorithm. Then, the ranking-based mutation operator is incorporated into five highly-competitive DE variants to solve the PEM fuel cell model parameter identification problems. The main contributions of this work are the proposed ranking-based DE variants and their application to the parameter identification problems of PEM fuel cell models. Experiments have been conducted by using both the simulated voltage–current data and the data obtained from the literature to validate the performance of our approach. The results indicate that the ranking-based DE methods provide better results with respect to the solution quality, the convergence rate, and the success rate compared with their corresponding original DE methods. In addition, the voltage–current characteristics obtained by our approach are in good agreement with the original voltage–current curves in all cases. - Highlights: • A simple and generic ranking-based mutation operator is presented in this paper. • Several DE (differential evolution) variants are used to solve the parameter identification of PEMFC (proton exchange membrane fuel cells) model. • Results show that our method accelerates the process of parameter identification. • The V–I characteristics are in very good agreement with experimental data

Fuel reliability experience in Finland

International Nuclear Information System (INIS)

Kekkonen, L.

2015-01-01

Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

Modal testing and identification of a PWR fuel assembly

International Nuclear Information System (INIS)

Pisapia, S.; Collard, B.; Mori, V.; Bellizzi, S.

2003-01-01

This study aims at characterizing the vibratory behavior of a full-scale fuel assembly using an experimental approach. The effect of the assembly environment (air, stagnant water, and water under flow) is studied. The analysis of the test series shows that the vibratory behavior of full-scale fuel assembly is strongly nonlinear. An identification phase, based on temporal mean square criterion, allows us to obtain a nonlinear model representative of the first vibration mode of a fuel assembly. The selected class of models including damping and stiffness nonlinear terms is efficient in air, in stagnant water, and in water under flow. In all environments, the stiffness decreases with the displacement level and the damping increases with the velocity level. In the presence of water, the damping goes up and increases again with flowrate. (author)

Reliability of the fuel identification procedure used by COGEMA during cask loading for shipment to LA HAGUE

International Nuclear Information System (INIS)

Pretesacque, P.; Eid, M.; Zachar, M.

1993-01-01

This study has been carried out to demonstrate the reliability of the system of the spent fuel identification used by COGEMA and NTL prior to shipment to the reprocessing plant of La Hague. This was a prerequisite for the French competent authority to accept the 'burnup credit' assumption in the criticality assessment of spent fuel packages. The probability to load a non-irradiated and non-specified fuel assembly was considered as acceptable if our identification and irradiation status measurement procedures were used. Furthermore, the task analysis enabled us to improve the working conditions at reactor sites, the quality of the working documentation, and consequently to improve the reliability of the system. The NTL experience of transporting to La Hague, as consignor, more than 10,000 fuel assemblies since the date of implementation of our system in 1984 without any non-conformance on fuel identification, validated the formalism of this study as well as our assumptions on basic events probabilities. (J.P.N.)

Fire-Side Corrosion: A Case Study of Failed Tubes of a Fossil Fuel Boiler

Directory of Open Access Journals (Sweden)

Majid Asnavandi

2017-01-01

Full Text Available The failures of superheater and reheater boiler tubes operating in a power plant utilizing natural gas or mazut as a fuel have been analysed and the fire-side corrosion has been suggested as the main reason for the failure in boiler tubes. The tubes have been provided by a fossil fuel power plant in Iran and optical and electron microscopy investigations have been performed on the tubes as well as the corrosion products on their surfaces. The results showed that the thickness of the failed tubes is not uniform which suggests that fire-side corrosion has happened on the tubes. Fire-side corrosion is caused by the reaction of combustion products with oxide layers on the tube surface resulting in metal loss and consequently tubes fracture. However, the tubes corrosion behaviour did not follow the conventional models of the fire-side corrosion. Given that, using the corrosion monitoring techniques for these boiler tubes seems essential. As a result, the thickness of the boiler tubes in different parts of the boiler has been recorded and critical points are selected accordingly. Such critical points are selected for installation of corrosion monitoring probes.

Failed fuel detection device

International Nuclear Information System (INIS)

Kasahara, Yoshiyuki; Soroi, Masatoshi.

1992-01-01

A pair of coil springs each of different spring rigidity are disposed independently to an interface mechanism which engages a reactor core fuel assembly. The springing reaction of the coil springs is utilized for providing a structure capable of detaching. A driving portion vertically movable in an inner cylinder of a system main body interlocking with the intrface mechanism is disposed, as well as a system separation mechanism is disposed for conducting electromotive remote control when it is required. With such a constitution, although it has been necessary so far that a plurality of operators access the reactor core upper mechanisms, it is not necessary according to the device of the present invention. Accordingly, the problem of operator's exposure can be overcome. (I.S.)

Failed fuel rod detection method by ultrasonic wave

International Nuclear Information System (INIS)

Takamatsu, Masatoshi; Muraoka, Shoichi; Ono, Yukio; Yasojima, Yujiro.

1990-01-01

Ultrasonic wave signals sent from an ultrasonic receiving element are supplied to an evaluation circuit by way of a gate. A table for gate opening and closing timings at the detecting position in each of the fuel rods in a fuel assembly is stored in a memory. A fuel rod is placed between an ultrasonic transmitting element and the receiving element to determine the positions of the transmitting element and the receiving element by positional sensors. The opening and closing timings at the positions corresponding to the result of the detection are read out from the table, and the gates are opened and closed by the timing. This can introduce the ultrasonic wave signals transmitted through a control rod always to the evaluation circuit passing through the gate. Accordingly, the state of failure of the fuel rod can be detected accurately. (I.N.)

Repairing fuel for reinsertion

International Nuclear Information System (INIS)

Krukshenk, A.

1986-01-01

Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

Identification of accelerants, fuels and post-combustion residues using a colorimetric sensor array.

Science.gov (United States)

Li, Zheng; Jang, Minseok; Askim, Jon R; Suslick, Kenneth S

2015-09-07

A linear (1 × 36) colorimetric sensor array has been integrated with a pre-oxidation technique for detection and identification of a variety of fuels and post-combustion residues. The pre-oxidation method permits the conversion of fuel vapor into more detectable species and therefore greatly enhances the sensitivity of the sensor array. The pre-oxidation technique used a packed tube of chromic acid on an oxide support and was optimized in terms of the support and concentration. Excellent batch to batch reproducibility was observed for preparation and use of the disposable pre-oxidation tubes. Twenty automotive fuels including gasolines and diesel from five gasoline retailers were individually identifiable with no confusions or misclassifications in quintuplicate trials. Limits of detection were at sub-ppm concentrations for gasoline and diesel fuels. In addition, burning tests were performed on commonly used fire accelerants, and clear differentiation was achieved among both the fuels themselves and their volatile residues after burning.

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1990-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous radionuclides. A modeling approach is outlined and sample calculations are provided that show the effects on release due to a limited perforation area, to decreasing temperature, and to the partial occlusion of the perforated area by corrosion products. 8 refs., 2 tabs

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1989-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous 14 CO 2 . A modeling approach is outlined and sample calculations are provided that show the effect on release due to a limited perforation area, to decreasing temperature, and to partial occlusion of the perforated area by corrosion products. 5 refs., 4 tabs

Nuclear reactor internals construction and failed fuel rod detection system

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

Science.gov (United States)

Gross, Kenny C.

1994-01-01

Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

International Nuclear Information System (INIS)

Gross, K.C.

1994-01-01

Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as ''background'' gases, further reducing the number of trial node combinations. Lastly, a ''fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs

Improvement of failed fuel detection system of light water reactor

International Nuclear Information System (INIS)

Chung, M.K.; Kang, H.D.; Cho, S.W.; Lee, K.W.

1981-01-01

Multi-task DAAS system by utilizing PDP-11/23 computer was assembled and tested for its performances. By connecting four Ge(Li) detectors to this DAAS, test experiments were done to prove system capability for detection and analysis of any fission gases resolved in four independently sampled primary cooling water from a power reactor. Appropriate computer programs were also introduced for this application and satisfactory results were obtained. Further application of this DAAS to the quality test of fuel pins (uniform distribution of enriched uranium in fresh fuel pellets), a prototype fuel scanner system was designed, constructed and tested. Operational principle of this system is based on the determination of 235 U/ 238 U abundance ratio in pellets by precision spectrometry or gamma-rays which are emitted from a portion of fuel pellets. For the uniform scanning, rotational and traverse motions at pre-selected speeds were applied to a fuel pin under tests. A long lens magnetic beta-spectrometer of Argonne National Laboratory was transferred to KAERI and re-installed for future precision beta-gamma spectroscopic research works on short-lived fission products nuclei

Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

International Nuclear Information System (INIS)

Eid, M.; Zachar, M.; Pretesacque, P.

1991-01-01

The Spent Fuel Identification for Flask Loading (SFIFL) procedure designed by COGEMA is analysed and its reliability calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states) is used for calculations. A maximal cut contains more than one failure which can split into two cuts (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

Reliability of the spent fuel identification for flask loading procedure used by COGEMA for fuel transport to La Hague

International Nuclear Information System (INIS)

Eid, M.; Zachar, M.; Pretesacque, P.

1990-01-01

The Spent Fuel Identification for Flask Loading, SFIFL, procedure designed by COGEMA is analysed and its reliability is calculated. The reliability of the procedure is defined as the probability of transporting only approved fuel elements for a given number of shipments. The procedure describes a non-coherent system. A non-coherent system is the one in which two successive failures could result in a success, from the system mission point of view. A technique that describes the system with the help of its maximal cuts (states), is used for calculations. A maximal cut contains more than one failure can split into two cuts, (sub-states). Cuts splitting will enable us to analyse, in a systematic way, non-coherent systems with independent basic components. (author)

Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

International Nuclear Information System (INIS)

Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

2002-01-01

Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

Energy Technology Data Exchange (ETDEWEB)

Patrick, W.C.; Carlson, R.C.; Rector, N.L.

1981-11-30

The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date.

Instrumentation Report No. 2: identification, evaluation, and remedial actions related to transducer failures at the spent fuel test-climax

International Nuclear Information System (INIS)

Patrick, W.C.; Carlson, R.C.; Rector, N.L.

1981-01-01

The Spent Fuel Test-Climax (SFT-C) is a test of the feasibility of safe and reliable short-term storage and retrieval of spent fuel from commercial nuclear reactors. In support of operational and technical goals of the test, about 850 channels of instrumentation have been installed at the SFT-C. Failure of several near-field instruments began less than six months after emplacement of 11 canisters of spent fuel and activation of six thermally similar simulators. The failed units were linear potentiometers (used to make displacement measurements) and vibrating wire stressmeters (used to make change-in-stress measurements). This report discusses the observed problems and remedial actions taken to date

Gas release from a failed fuel pin after reactor shut-down

International Nuclear Information System (INIS)

Pshenichnikov, B.V.

1975-01-01

A mathematical model of gassing from a hypothetical core fuel element in the active zone of a stopped water-moderated reactor was analysed to investigate the process of liberation of gaseous fission products from an unpressurized fuel element. A one-dimensional problem was obtained as a result of the accepted hypotheses. A fault was assumed to have occured during reactor operation; at the same time, a vapour-gas mixture was considered to be present under the envelope at reactor working pressure by the moment of stoppage. An approximative estimation was made of the retardation time of pressure balancing at the open end of the fuel element, and also of the amount of total gas remaining in the gap under the fuel element envelope after pressure drop in the reactor. Estimation of retardation time permitted to conclude that pressure in the nonhermetic fuel element envelope follows pressure fluctuation in the reactor in the course of cooling, the retardation time of pressure balancing outside and inside the fuel element lasting but a few seconds

KNF's fuel service technologies and experiences

International Nuclear Information System (INIS)

Shin, Jung Cheol; Kwon, Jung Tack; Kim, Jaeik; Park, Jong Youl; Kim, Yong Chan

2009-01-01

In Korea, since 1978, the commercial nuclear power plant was operated. After 10 years, from 1988, the nuclear fuel was produced by KNF (Korea Nuclear Fuel). The Fuel Service Team was established at KNF in 1995. Through the technical self reliance periods in cooperate with advanced foreign companies for 5 years, KNF has started to carry out fuel service activities onsite in domestic nuclear power plants. By ceaseless improving and advancing our own methodologies, after that, KNF is able to provide the most safe and reliable fuel repair services and poolside examinations including the root cause analysis of failed fuels. Recently, KNF developed the fuel cleaning system using ultrasonic technique for crud removal, and the CANDU fuel sipping system to detect a failed fuel bundle in PHWR. In this paper, all of KNF's fuel service technologies are briefly described, and the gained experience in shown

Identification of potential safety-related incidents applicable to a breeder fuel reprocessing plant

International Nuclear Information System (INIS)

Perkins, W.C.

1980-01-01

The current emphasis on safety in all phases of the nuclear fuel cycle requires that safety features be identified and included in designs of nuclear facilities at the earliest possible stage. A popular method for the early identification of these safety features is the Preliminary Hazards Analysis. An extension of this analysis is to illustrate the nature of a hazard by its effects in accident situations, that is, to identify what are called safety-related incidents. Some useful tools are described which have been used at the Savannah River Laboratory, SRL, to make Preliminary Hazards Analyses as well as safety analyses of facilities for processing spent nuclear fuels from both power and production reactors. These tools have also been used in safety studies of waste handling operations at the Savannah River Plant. The tools are the SRL Incidents Data Bank and the What If meeting. The application of this methodology to a proposed facility which has breeder fuel reprocessing capability, the Hot Experimental Facility (HEF) is illustrated

Prescribed burning in ponderosa pine: fuel reductions and redistributing fuels near boles to prevent injury

Science.gov (United States)

Prescribed burning can be an effective tool for thinning forests and reducing fuels to lessen wildfire risks. However, prescribed burning sometimes fails to substantially reduce fuels and sometimes damages/kills valuable, large trees. This study compared fuel reductions between fall and spring pre...

Identification of a Methane Oxidation Intermediate on Solid Oxide Fuel Cell Anode Surfaces with Fourier Transform Infrared Emission.

Science.gov (United States)

Pomfret, Michael B; Steinhurst, Daniel A; Owrutsky, Jeffrey C

2013-04-18

Fuel interactions on solid oxide fuel cell (SOFC) anodes are studied with in situ Fourier transform infrared emission spectroscopy (FTIRES). SOFCs are operated at 800 °C with CH4 as a representative hydrocarbon fuel. IR signatures of gas-phase oxidation products, CO2(g) and CO(g), are observed while cells are under load. A broad feature at 2295 cm(-1) is assigned to CO2 adsorbed on Ni as a CH4 oxidation intermediate during cell operation and while carbon deposits are electrochemically oxidized after CH4 operation. Electrochemical control provides confirmation of the assignment of adsorbed CO2. FTIRES has been demonstrated as a viable technique for the identification of fuel oxidation intermediates and products in working SOFCs, allowing for the elucidation of the mechanisms of fuel chemistry.

Failing Failed States

DEFF Research Database (Denmark)

Holm, Hans-Henrik

2002-01-01

coverage. A Danish survey of newsrooms shows that the national world-view and prevalent news criteria prevent consistent coverage. It is argued that politicians are the ones who determine national agendas: it is from political initiatives, rather than media coverage, that failing states and humanitarian......When states are failing, when basic state functions are no longer carried out, and when people have no security, humanitarian crises erupt. In confronting this problem, the stronger states have followed an ad hoc policy of intervention and aid. In some cases, humanitarian disasters have resulted...... from inaction. Often, the media are blamed. Politicians complain about the media when they interfere (the CNN effect), and when they do not. This article looks at how the media do cover failing states. Sierra Leone and Congo are used as examples. The analysis shows that there is little independent...

Simulation of leaking fuel rods

International Nuclear Information System (INIS)

Hozer, Z.

2006-01-01

The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

KNF's fuel service technologies and experiences

Energy Technology Data Exchange (ETDEWEB)

Shin, Jung Cheol; Kwon, Jung Tack; Kim, Jaeik; Park, Jong Youl; Kim, Yong Chan [KNF, Daejeon (Korea, Republic of)

2009-04-15

In Korea, since 1978, the commercial nuclear power plant was operated. After 10 years, from 1988, the nuclear fuel was produced by KNF (Korea Nuclear Fuel). The Fuel Service Team was established at KNF in 1995. Through the technical self reliance periods in cooperate with advanced foreign companies for 5 years, KNF has started to carry out fuel service activities onsite in domestic nuclear power plants. By ceaseless improving and advancing our own methodologies, after that, KNF is able to provide the most safe and reliable fuel repair services and poolside examinations including the root cause analysis of failed fuels. Recently, KNF developed the fuel cleaning system using ultrasonic technique for crud removal, and the CANDU fuel sipping system to detect a failed fuel bundle in PHWR. In this paper, all of KNF's fuel service technologies are briefly described, and the gained experience in shown.

A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

International Nuclear Information System (INIS)

Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya

2001-05-01

In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10 25 m -2 , respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

Reactor fuel exchanging facility

International Nuclear Information System (INIS)

Kubota, Shin-ichi.

1981-01-01

Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

Simplified fuel cell system model identification

Energy Technology Data Exchange (ETDEWEB)

Caux, S.; Fadel, M. [Laboratoire d' Electrotechnique et d' Electronique Industrielle, Toulouse (France); Hankache, W. [Laboratoire d' Electrotechnique et d' Electronique Industrielle, Toulouse (France)]|[Laboratoire de recherche en Electronique, Electrotechnique et Systemes, Belfort (France); Hissel, D. [Laboratoire de recherche en Electronique, Electrotechnique et Systemes, Belfort (France)

2006-07-01

This paper discussed a simplified physical fuel cell model used to study fuel cell and supercap energy applications for vehicles. Anode, cathode, membrane, and electrode elements of the cell were modelled. A quasi-static Amphlett model was used to predict voltage responses of the fuel cell as a function of the current, temperature, and partial pressures of the reactive gases. The potential of each cell was multiplied by the number of cells in order to model a fuel cell stack. The model was used to describe the main phenomena associated with current voltage behaviour. Data were then compared with data from laboratory tests conducted on a 20 cell stack subjected to a current and time profile developed using speed data from a vehicle operating in an urban environment. The validated model was used to develop iterative optimization algorithms for an energy management strategy that linked 3 voltage sources with fuel cell parameters. It was concluded that classic state and dynamic measurements using a simple least square algorithm can be used to identify the most important parameters for optimal fuel cell operation. 9 refs., 1 tab., 6 figs.

Nuclear fuel shipping inspection device

International Nuclear Information System (INIS)

Takahashi, Toshio; Hada, Koji.

1988-01-01

Purpose: To provide an nuclear fuel shipping inspection device having a high detection sensitivity and capable of obtaining highly reliable inspection results. Constitution: The present invention concerns a device for distinguishing a fuel assembly having failed fuel rods in LMFBR type reactors. Coolants in a fuel assembly to be inspected are collected by a sampling pipeway and transferred to a filter device. In the filter device, granular radioactive corrosion products (CP) in the coolants are captured, to reduce the background. The coolants, after being passed through the filter device, are transferred to an FP catching device and gamma-rays of iodine and cesium nuclides are measured in FP radiation measuring device. Subsequently, the coolants transferred to a degasing device to separate rare gas FP in the coolants from the liquid phase. In a case if rare gas fission products are detected by the radiation detector, it means that there is a failed fuel rod in the fuel assembly to be inspected. Since the CP and the soluble FP are separated and extracted for the radioactivity measurement, the reliability can be improved. (Kamimura, M.)

Fuel assemblies for BWR type reactors

International Nuclear Information System (INIS)

Ishizuka, Takao.

1981-01-01

Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

Intrauterine adhesions as a risk factor for failed first-trimester pregnancy termination.

Science.gov (United States)

Luk, Janelle; Allen, Rebecca H; Schantz-Dunn, Julianna; Goldberg, Alisa B

2007-10-01

Risk factors for failed first-trimester surgical abortion include endometrial distortion caused by leiomyomas, uterine anomalies and malposition and cervical stenosis. This report introduces intrauterine adhesions as an additional risk factor. A multiparous woman presented for pregnancy termination at 6 weeks' gestation. Three suction-curettage attempts failed to remove what appeared to be an intrauterine pregnancy. Rising beta-hCG levels and concern for an interstitial ectopic pregnancy prompted a diagnostic laparoscopy and exploratory laparotomy without the identification of an ectopic pregnancy. After methotrexate treatment failed, the patient underwent ultrasound-guided hysteroscopy and suction curettage using a cannula with a whistle-cut aperture for the successful removal of a pregnancy implanted behind intrauterine adhesions. Intrauterine adhesions are a cause of failed surgical abortion. Ultrasound-guided hysteroscopy may be required for diagnosis.

Fuel assembly identification by magnetic scanning

International Nuclear Information System (INIS)

Badurek, G.

1986-09-01

In order to identify individual fuel assemblies by a magnetic fingerprint, investigations were made on iron inclusions in fuel elements and a method was developed to measure these by magnetically scanning the element. The fuel assembly is drawn with constant speed through a homogeneous magnetic field to magnetize iron inclusions. Resulting inhomogeneous magnetic dipole fields induce a voltage difference in pick up coils which is proportional to the mass of the inclusion. Using lock-in technique 3 mg pieces of steel wire on the surface of the fuel element were detected while the lower limit for the center of an assembly for ferromagnetic spheres was 50 mg. In single rods ferromagnetic samples of 1 mg were detected regardless of geometric form or location. With minor modifications of the measuring procedure the sensitivity limit can be improved to about 10 mg at the center of an assembly. In the KWU-fuel at Zwentendorf no iron inclusions were found

Microbiological and molecular identification of bacterial species isolated from nasal and oropharyngeal mucosa of fuel workers in Riyadh, Saudi Arabia.

Science.gov (United States)

AlWakeel, Suaad S

2017-09-01

This study aimed to determine the bacterial species colonizing the nasal and oropharyngeal mucosa of fuel workers in Central Riyadh, Saudi Arabia on a microbiological and molecular level. Throat and nasal swab samples were obtained from 29 fuel station attendants in the period of time extending from March to May 2014 in Riyadh, Saudi Arabia. Microbiological identification techniques were utilized to identify the bacterial species isolated. Antibiotic sensitivity was assessed for each of the bacterial isolates. Molecular identification techniques based on PCR analysis of specific genomic sequences was conducted and was the basis on which phylogeny representation was done for 10 randomly selected samples of the isolates. Blood was drawn and a complete blood count was conducted to note the hematological indices for each of the study participants. Nineteen bacterial species were isolated from both the nasal cavity and the oropharynx including Streptococcus thoraltensis , alpha-hemolytic streptococci, Staphylococcus hominis , coagulase-negative staphylococci, Leuconostoc mesenteroides , Erysipelothrix rhusiopathiae and several others. We found 100% sensitivity of the isolates to ciprofloxacin, cefuroxime and gentamicin. Whereas cefotaxime and azithromycin posted sensitivities of 85.7% and 91.4%, respectively. Low sensitivities (fuel products may be a contributing factor to bacterial colonization of the respiratory tract in fuel workers.

Some factors to consider in handling and storing spent fuel

International Nuclear Information System (INIS)

Bailey, W.J.

1985-11-01

This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

Fuel design and operational experience in Loviisa NPP, future trends in fuel issues

International Nuclear Information System (INIS)

Terasvirta, R.

2001-01-01

This paper summarizes the past operational experience of nuclear fuel with reference to most significant design changes during the years. In general, the fuel behaviour in Loviisa NPP in terms of leaking fuel assemblies has been good. The major improvements by fuel design changes in Lovissa NPP, including rod elongation margin, change in the pellet design and manufacturing process, upper grid modifications, change of material in the spacer grids and reduction of the shroud tube thickness are discussed and related to the number of failed fuel assemblies. The detailed investigation of fuel failure rates as function of different fuel and operation characteristics allows to classify the leaking fuel assemblies according to the cause of failure. In a brief discussion concerning new changes in the safety guide for nuclear design limits, re-issued by the Finnish Safety Authority (STUK), the frequencies for class 1 and class 2 accidents are determined. Another change in this guide is the introduction of design limits for the number of fuel rods experiencing DNB in class 1 accidents and number of failed rods in class 2 accidents. It is concluded that as far as normal operation is concerned, there seems to be sufficiently large margin between present operational limits in Loviisa and the design limits. The real limits do not come from fuel behaviour in the normal operation or operational occurrences but from the accident behaviour. At the moment, fuel assembly burnup extension beyond 45 MWd/kgU is clearly out of the question before further information and positive results are obtained on high burnup fuel behaviour in accident conditions

Studies and research concerning BNFP: evaluation of spent-fuel-examination techniques for the Barnwell Nuclear Fuel Plant

International Nuclear Information System (INIS)

Anderson, R.T.; Gray, J.H.; Rogell, M.L.

1982-09-01

A study was made of various examinations which could be remotely performed on a production basis with spent fuel at the Barnwell Nuclear Fuel Plant (BNFP). These techniques could form an integral portion of fuel disassembly and canning operations. Their benefits accrue to either improved fuel storage, reprocessing, or both. In conjunctoin with these studies, evaluations have been made of the operational impact of receiving failed or canned fuel at the BNFP

Fuel element services

International Nuclear Information System (INIS)

Marta, H.; Alvarez, P.; Jimenez, J.

2006-01-01

Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

Microbiological and molecular identification of bacterial species isolated from nasal and oropharyngeal mucosa of fuel workers in Riyadh,

Directory of Open Access Journals (Sweden)

Suaad S. AlWakeel

2017-09-01

Full Text Available This study aimed to determine the bacterial species colonizing the nasal and oropharyngeal mucosa of fuel workers in Central Riyadh, Saudi Arabia on a microbiological and molecular level. Throat and nasal swab samples were obtained from 29 fuel station attendants in the period of time extending from March to May 2014 in Riyadh, Saudi Arabia. Microbiological identification techniques were utilized to identify the bacterial species isolated. Antibiotic sensitivity was assessed for each of the bacterial isolates. Molecular identification techniques based on PCR analysis of specific genomic sequences was conducted and was the basis on which phylogeny representation was done for 10 randomly selected samples of the isolates. Blood was drawn and a complete blood count was conducted to note the hematological indices for each of the study participants. Nineteen bacterial species were isolated from both the nasal cavity and the oropharynx including Streptococcus thoraltensis, alpha-hemolytic streptococci, Staphylococcus hominis, coagulase-negative staphylococci, Leuconostoc mesenteroides, Erysipelothrix rhusiopathiae and several others. We found 100% sensitivity of the isolates to ciprofloxacin, cefuroxime and gentamicin. Whereas cefotaxime and azithromycin posted sensitivities of 85.7% and 91.4%, respectively. Low sensitivities (<60% sensitivity to the antibiotics ampicillin, erythromycin, clarithromycin and norfloxacin were observed. Ninety-seven percent similarity to the microbial bank species was noted when the isolates were compared to it. Most hematological indices recorded were within the normal range. In conclusion, exposure to toxic fumes and compounds within fuel products may be a contributing factor to bacterial colonization of the respiratory tract in fuel workers.

TMI-2 spent fuel shipping

International Nuclear Information System (INIS)

Quinn, G.J.; Burton, H.M.

1985-01-01

TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

A model for pressure in an LMFBR duct due to discharge of gas from a failed fuel pin

International Nuclear Information System (INIS)

Srinivas, S.; Chopra, P.S.

1977-01-01

In this paper an analytical model for the calculation of pressure pulses in hexagonal ducts due to discharge of gas from a failed fuel pin is developed. The analysis yields the time history of the pressure pulse which can be used in the calculation of permanent deformation of the duct or in the assessment of the susceptibility of the duct to fracture. The real physical situation of gas discharging through a pin in a duct filled with liquid is complex to model. Here the phenomenon is modeled based on some reasonable assumptions. In this model the analysis is divided into two stages. In the first stage the gas expands as a spherical bubble, but the influence of the duct wall is taken into account. At the end of the first stage the spherical shape of the bubble is assumed to be lost and the gas is assumed to expand axially as a column. The analysis involves solving the continuity and momentum equations for the liquid along with the energy balance equation for the gas

Automatic fuel exchanging device

International Nuclear Information System (INIS)

Takahashi, Fuminobu.

1984-01-01

Purpose: To enable to designate the identification number of a fuel assembly in a nuclear reactor pressure vessel thereby surely exchanging the designated assembly within a short time. Constitution: Identification number (or letter) pressed on a grip of a fuel assembly is to be detected by a two-dimensional ultrasonic probe of a pull-up mechanism. When the detected number corresponds with the designated number, a control signal is outputted, whereby the pull-up drive control mechanism or pull-up mechanism responds to pull-up and exchange the fuel assembly of the identified number. With such a constitution, the fuel assembly can rapidly and surely be recognized even if pressed letters deviate to the left or right of the probe, and further, the hinge portion and the signal processing portion can be simplified. (Horiuchi, T.)

TREAT hodoscope interpretation. II. Fuel-state identification

International Nuclear Information System (INIS)

Wu, R.M.; Omberg, R.P.; Albrecht, R.W.

1982-01-01

By using the autoregressive-integrated-moving-average (ARIMA) process, the onset of fuel disposal and the restructured fuel states of a TREAT test can be unambiguously identified. The results of the ARIMA analyses on the TREAT L7 hodoscope data show the most probable time of the restructuring began at 14.038 seconds, and four restructured fuel states are required to interpret adequately the L7 hodoscope data

Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

International Nuclear Information System (INIS)

Arthur, S.E.; McNeish, J.

2002-01-01

The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

Modeling of coated fuel particles irradiation behavior

International Nuclear Information System (INIS)

Liang Tongxiang; Phelip, M.

2006-01-01

In this report, PANAMA code was used to estimate the CP performance under normal and accident condition. Under the normal irradiation test (1000 degree C 625 efpd, 10% FIMA), for intact CP fuel, failure fraction is in the level of 10 -7 . As-fabricated SiC failed particles results in the through coatings failed particles much earlier than the intact particles does, OPyC layer does not fail immediately after irradiation starts. The significant failures start at beyond the burnup of about 7% FIMA. Under the accident condition, the calculated results showed that when the heating temperature is much higher than 1850 degree C, the failure fraction of coated particle can reach the level of 1 percent. The CP fuel fails significantly if it has a buffer layer thinner than 65 urn, SiC layer thinner than 30 μm. High burnup CP need to develop small size kernel, thick buffer layer and thick SiC layer. (authors)

Fuel performance evaluation through iodine activity monitoring

International Nuclear Information System (INIS)

Anantharaman, K.; Chandra, R.

1995-01-01

The objective of the failed fuel detection system is to keep a watch on fuel behaviour during operation. This paper describes the evaluation of fuel behaviour by monitoring the activities of various isotopes of iodine both during steady state and during a reactor shutdown. The limitations of this approach also has been explained. The monitoring of tramp uranium for different types of release, namely fixed contamination and continuous release from fuel, is also presented. (author)

LMFBR fuel-design environment for endurance testing, primarily of oxide fuel elements with local faults

International Nuclear Information System (INIS)

Warinner, D.K.

1980-01-01

The US Department of Energy LMFBR Lines-of-Assurance are briefly stated and local faults are given perspective with an historical review and definition to help define the constraints of LMFBR fuel-element designs. Local-fault-propagation (fuel-element failure-propagation and blockage propagation) perceptions are reviewed. Fuel pin designs and major LMFBR parameters affecting pin performance are summarized. The interpretation of failed-fuel data is aided by a discussion of the effects of nonprototypicalities. The fuel-pin endurance expected in the US, USSR, France, UK, Japan, and West Germany is outlined. Finally, fuel-failure detection and location by delayed-neutron and gaseous-fission-product monitors are briefly discussed to better realize the operational limits

Management number identification method for nuclear fuel assembly

International Nuclear Information System (INIS)

Furuya, Nobuo; Mori, Kazuma.

1995-01-01

In the present invention, a management number indicated to appropriate portions of a fuel assembly can be read with no error for the management of nuclear fuel materials in the nuclear fuel assembly (counting management) and physical protection: PP. Namely, bar codes as a management number are printed by electrolytic polishing to one or more portions of a side surface of an upper nozzle of the assembly, an upper surface of a clamp and a side surface of a lower nozzle. The bar codes are read by a reader at one or more portions in a transporting path for transporting the fuel assembly and at a fuel detection device disposed in a fuel storage pool. The read signals are inputted to a computer. With such procedures, the nuclear fuel assembly can be identified with no error by reading the bar codes and without applying no danger to a human body. Since the reader is disposed in the course of the transportation and test for the assembly, and the read signals are inputted to the computer, the management for the counting number and PP is facilitated. (I.S.)

Best Practices for Fuel System Contamination Detection and Remediation

Science.gov (United States)

2016-01-15

The University of Dayton Research Institute Best Practices for Fuel System Contamination Detection and Remediation Final Report Marlin D... Remediation Executive Summary: Fuel contamination is a broad term commonly applied to anything that causes a fuel test to fail quality assurance...Statement A: Approved for public release: distribution unlimited. 1 Best Practices for Fuel System Contamination Detection and Remediation Contents

Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

International Nuclear Information System (INIS)

Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

1995-01-01

The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described

Advances in HTGR fuel performance models

International Nuclear Information System (INIS)

Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

1985-01-01

Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

Linear identification and model adjustment of a PEM fuel cell stack

Energy Technology Data Exchange (ETDEWEB)

Kunusch, C; Puleston, P F; More, J J [LEICI, Departamento de Electrotecnia, Universidad Nacional de La Plata, calle 1 esq. 47 s/n, 1900 La Plata (Argentina); Consejo de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Husar, A [Institut de Robotica i Informatica Industrial (CSIC-UPC), c/ Llorens i Artigas 4-6, 08028 Barcelona (Spain); Mayosky, M A [LEICI, Departamento de Electrotecnia, Universidad Nacional de La Plata, calle 1 esq. 47 s/n, 1900 La Plata (Argentina); Comision de Investigaciones Cientificas (CIC), Provincia de Buenos Aires (Argentina)

2008-07-15

In the context of fuel cell stack control a mayor challenge is modeling the interdependence of various complex subsystem dynamics. In many cases, the states interaction is usually modeled through several look-up tables, decision blocks and piecewise continuous functions. Many internal variables are inaccessible for measurement and cannot be used in control algorithms. To make significant contributions in this area, it is necessary to develop reliable models for control and design purposes. In this paper, a linear model based on experimental identification of a 7-cell stack was developed. The procedure followed to obtain a linear model of the system consisted in performing spectroscopy tests of four different single-input single-output subsystems. The considered inputs for the tests were the stack current and the cathode oxygen flow rate, while the measured outputs were the stack voltage and the cathode total pressure. The resulting model can be used either for model-based control design or for on-line analysis and errors detection. (author)

Fuel-pin cladding transient failure strain criterion

International Nuclear Information System (INIS)

Bard, F.E.; Duncan, D.R.; Hunter, C.W.

1983-01-01

A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test (TREAT) facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion

Physics operating experience and fuel management of RAPS-1

International Nuclear Information System (INIS)

Nakra, A.N.; Purandare, H.D.; Srinivasan, K.R.; Rastogi, B.P.

1976-01-01

Rajasthan Atomic Power Station Unit-1 achieved criticality on August 11, 1972. Thereafter the reactor was brought to power, in November, 1972. Due to non-availability of the depleted fuel, the loading of which was necessary to obtain full power to begin with, the core was loaded with all natural uranium fuel and only 70% of the full power could be achieved. During the reactor operation for the last three years, the reactor has seen more than one effective full power year and about 1400 fresh fuel bundles have been loaded in the core. The reactor was subjected to about 150 power cycles resulting in more than 30% variation in operating power level and about 10 fuel bundles have failed. For satisfactory fuel management and refuelling decisions, a three dimensional simulator TRIVENI was developed. This was extensively tested during the start-up experiments and was found to be a satisfactory tool for day to day operation of the plant. In this paper, a brief account of analysis of the start-up experiments, approach to full power, power distortions and flux peaking, fuel management service and analysis of the failed fuel data has been given. (author)

Checklist for transition to new highway fuel(s).

Energy Technology Data Exchange (ETDEWEB)

Risch, C.; Santini, D.J. (Energy Systems)

2011-12-15

Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s) of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.

Telescope sipping the optimum fuel leak detection system

International Nuclear Information System (INIS)

Deleryd, R.

1998-01-01

The TELESCOPE Sipping technology is an evolutionary development from previous ABB fuel leak systems used in LWR reactors. The system utilizes the existing dynamics that cause numerous fission products to leak from a failed fuel rod when the fuel assembly is raised from a reactor core during core fuel alterations. The system can also be used by repair work in pool side inspection in order to detect leaking rods or to verify reconstituted assemblies as non leakers. (author)

Contribution to the study on the contamination of the primary coolant of pressurized water reactors by fission products. Application for evaluation of failed fuel elements

International Nuclear Information System (INIS)

Bourgeois, Pierre.

1979-01-01

After describing the radioactive substances present in the primary system, the consequences of this contamination are studied. The many publications relating to the emergence of fission products from the fuel pins are then analyzed. This study reaches the conclusion that none of the proposed methods makes it possible to interpret the activity determinations made in power reactors. This is why the opposite approach is adopted consisting in suggesting a model as from in-reactor measurements. This model will apply in the conditions for which it has been adjusted, i.e. at a steady nominal power in a 900 MWe pressurized water reactor. The originality of this model lies in the determination of a probable emergence of the can-fuel set which reflects the state of the can and the demonstration of a trapping of the iodine in the set, the consequence of which is a lower release than that of the rare gases in the primary system. After comparing the experimental results available in the literature with the results of the model, it is shown how to proceed in order to evaluate the number of failed fuel elements. Finally, on the basis of the experiments performed at the Grenoble Nuclear Study Centre, the mean and maximal predictable activites are calculated for load following working. The study of the transients is completed by a more accurate calculation of the iodine releases when the reactor is shut down which confirms the main hypotheses of the steady state model [fr

Breached fuel location in FFTF by delayed neutron monitor triangulation

International Nuclear Information System (INIS)

Bunch, W.L.; Tang, E.L.

1985-10-01

The Fast Flux Test Facility (FFTF) features a three-loop, sodium-cooled 400 MWt mixed oxide fueled reactor designed for the irradiation testing of fuels and materials for use in liquid metal cooled fast reactors. To establish the ultimate capability of a particular fuel design and thereby generate information that will lead to improvements, many of the fuel irradiations are continued until a loss of cladding integrity (failure) occurs. When the cladding fails, fission gas escapes from the fuel pin and enters the reactor cover gas system. If the cladding failure permits the primary sodium to come in contact with the fuel, recoil fission products can enter the sodium. The presence of recoil fission products in the sodium can be detected by monitoring for the presence of delayed neutrons in the coolant. It is the present philosophy to not operate FFTF when a failure has occurred that permits fission fragments to enter the sodium. Thus, it is important that the identity and location of the fuel assembly that contains the failed cladding be established in order that it might be removed from the core. This report discusses method of location of fuel element when cladding is breached

Express diagnostics of WWER fuel rods at nuclear power plants

International Nuclear Information System (INIS)

Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

2009-01-01

Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

Fuel manufacturing and utilization

International Nuclear Information System (INIS)

2005-01-01

The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

Operational Experience of Nuclear Fuel in Finnish Nuclear Power Plants (with Emphasis on WWER Fuel)

International Nuclear Information System (INIS)

Teraesvirta, R.

2009-01-01

The four operating nuclear reactors in Finland, Loviisa-1 and -2 and Olkiluoto-1 and -2 have now operated approximately 30 years. The overall operational experience has been excellent. Load factors of all units have been for years among the highest in the world. The development of the fuel designs during the years has enabled remarkable improvement in the fuel performance in terms of burnup. Average discharge burnup has increased more than 30 percent in all Finnish reactor units. A systematic inspection of spent fuel assemblies, and especially all failed fuel assemblies, is a good and useful practise employed in Finland. A possibility to inspect the fuel on site using a pool side inspection facility is a relatively economic way to find out root causes of fuel failures and thereby facilitate developing remedies to prevent similar failures in the future

PWR fuel inspection and repair technology development in the Republic of Korea

International Nuclear Information System (INIS)

Park, J.Y.

1998-01-01

As of September 1997, 10 PWRs and 2 PHWRs generate 10,320MW electricity in Korea. And another 8 PWRs and 2 PHWRs will be constructed by 2006. These will need about 400 MTU of PWR fuels and 400 MTU of PHWR fuels. To improve average burnup, thermal power, fuel usability and plant safety, better poolside fuel service technologies are strongly recommended as well as the fuel design and fabrication technology improvements. During the last twenty years of nuclear power plant operation in Korea, more than 4,000 fuel assemblies has been used. At the site, continuous coolant activity measurement, pool-side visual inspection and ultrasonic tests have been performed. Some of the fuels are damaged or failed for various reasons. Some of the defected fuels were examined in hot cell to investigate the cause of failure. Even though 30 PWR fuel assemblies were repaired by foreign engineers, fuel inspection and repair technologies are not established yet. Various kind of design for the fuel make the inspection, repair and reconstitution equipment more complex. As a result, recently, a plant to obtain overall technology for poolside fuel inspection, failed fuel repair and reconstitution through R and D activities are set forth. (author)

Identification of vapor-phase chemical warfare agent simulants and rocket fuels using laser-induced breakdown spectroscopy

International Nuclear Information System (INIS)

Stearns, Jaime A.; McElman, Sarah E.; Dodd, James A.

2010-01-01

Application of laser-induced breakdown spectroscopy (LIBS) to the identification of security threats is a growing area of research. This work presents LIBS spectra of vapor-phase chemical warfare agent simulants and typical rocket fuels. A large dataset of spectra was acquired using a variety of gas mixtures and background pressures and processed using partial least squares analysis. The five compounds studied were identified with a 99% success rate by the best method. The temporal behavior of the emission lines as a function of chamber pressure and gas mixture was also investigated, revealing some interesting trends that merit further study.

Identification of vapor-phase chemical warfare agent simulants and rocket fuels using laser-induced breakdown spectroscopy

Energy Technology Data Exchange (ETDEWEB)

Stearns, Jaime A.; McElman, Sarah E.; Dodd, James A.

2010-05-01

Application of laser-induced breakdown spectroscopy (LIBS) to the identification of security threats is a growing area of research. This work presents LIBS spectra of vapor-phase chemical warfare agent simulants and typical rocket fuels. A large dataset of spectra was acquired using a variety of gas mixtures and background pressures and processed using partial least squares analysis. The five compounds studied were identified with a 99% success rate by the best method. The temporal behavior of the emission lines as a function of chamber pressure and gas mixture was also investigated, revealing some interesting trends that merit further study.

Determination and microscopic study of incipient defects in irradiated power reactor fuel rods. Final report

International Nuclear Information System (INIS)

Pasupathi, V.; Perrin, J.S.; Roberts, E.

1978-05-01

This report presents the results of nondestructive and destructive examinations carried out on the Point Beach-1 (PWR) and Dresden-3 (BWR) candidate fuel rods selected for the study of pellet-clad interaction (PCI) induced incipient defects. In addition, the report includes results of examination of sections from Oskarshamn-1 (BWR) fuel rods. Eddy current examination of Point Beach-1 rods showed indications of possible incipient defects in the fuel rods. The profilometry and the gamma scan data also indicated that the source of the eddy current indications may be incipient defects. No failed rods or rods with incipient failure were found in the sample from Point Beach-1. Despite the lack of success in finding incipient defects and filed rods, the mechanism for fuel rod failures in Point Beach-1 is postulated to be PCI-related, with high startup rates and fuel handling being the key elements. Nine out of the 10 candidate fuel rods from Dresden-3 (BWR) were failed, and all the failed rods had leaked water so that the initial mechanism was observed. Examination of clad inner surfaces of the specimens from failed and unfailed rods showed fuel deposits of widely varying appearance. The deposits were found to contain uranium, cesium, and tellurium. Transmission electron microscopy of clad specimens showed evidence of microscopic strain. Metallographic examination of fuel pellets from the peak transient power location showed extensive grain boundary separation and axial movement of the fuel indicative of rapid release of fission products. Examination of Oskarshamn clad specimens did not show any stress corrosion crack (SCC) type defects. The defects found in the examinations appear to be related to secondary hydriding. The clad inner surface of the Oskarshamn specimens also showed uranium-rich deposits of varying features

Fuel-in-air FY07 summary report

Energy Technology Data Exchange (ETDEWEB)

Hanson, Brady D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wittman, Richard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wu, Wesley [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, Rick W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2008-01-22

Results of the testing program to determine fractional release rates and particle size distributions from failed commercial spent fuel related to the operations in the surface facility at Yucca Mountain are presented.

Data-driven simultaneous fault diagnosis for solid oxide fuel cell system using multi-label pattern identification

Science.gov (United States)

Li, Shuanghong; Cao, Hongliang; Yang, Yupu

2018-02-01

Fault diagnosis is a key process for the reliability and safety of solid oxide fuel cell (SOFC) systems. However, it is difficult to rapidly and accurately identify faults for complicated SOFC systems, especially when simultaneous faults appear. In this research, a data-driven Multi-Label (ML) pattern identification approach is proposed to address the simultaneous fault diagnosis of SOFC systems. The framework of the simultaneous-fault diagnosis primarily includes two components: feature extraction and ML-SVM classifier. The simultaneous-fault diagnosis approach can be trained to diagnose simultaneous SOFC faults, such as fuel leakage, air leakage in different positions in the SOFC system, by just using simple training data sets consisting only single fault and not demanding simultaneous faults data. The experimental result shows the proposed framework can diagnose the simultaneous SOFC system faults with high accuracy requiring small number training data and low computational burden. In addition, Fault Inference Tree Analysis (FITA) is employed to identify the correlations among possible faults and their corresponding symptoms at the system component level.

Consequences of Fuel Failure on Criticality Safety of Used Nuclear Fuel

International Nuclear Information System (INIS)

Marshall, William J.; Wagner, John C.

2012-09-01

This report documents work performed for the Department of Energy's Office of Nuclear Energy (DOENE) Fuel Cycle Technologies Used Fuel Disposition Campaign to assess the impact of fuel reconfiguration due to fuel failure on the criticality safety of used nuclear fuel (UNF) in storage and transportation casks. This work was motivated by concerns related to the potential for fuel degradation during extended storage (ES) periods and transportation following ES, but has relevance to other potential causes of fuel reconfiguration. Commercial UNF in the United States is expected to remain in storage for longer periods than originally intended. Extended storage time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications for virtually all aspects of a UNF storage and transport system's performance. The potential impact of fuel reconfiguration on the safety of UNF in storage and transportation is dependent on the likelihood and extent of the fuel reconfiguration, which is not well understood and is currently an active area of research. The objective of this work is to assess and quantify the impact of postulated failed fuel configurations on the criticality safety of UNF in storage and transportation casks. Although this work is motivated by the potential for fuel degradation during ES periods and transportation following ES, it has relevance to fuel reconfiguration due to the effects of high burnup. Regardless of the ultimate disposition path, UNF will need to be transported at some point in the future. To investigate and quantify the impact of fuel reconfiguration on criticality safety limits, which are given in terms of the effective neutron multiplication factor, a set of failed fuel

Failure position detection device for nuclear fuel rod

International Nuclear Information System (INIS)

Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

1987-01-01

Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.)

Failure position detection device for nuclear fuel rod

Energy Technology Data Exchange (ETDEWEB)

Ishida, Takeshi; Higuchi, Shin-ichi; Ito, Masaru; Matsuda, Yasuhiko

1987-03-24

Purpose: To easily detect failure position of a nuclear fuel rod by relatively moving an air-tightly shielded detection portion to a fuel rod. Constitution: For detecting the failure position of a leaked fuel assembly, the fuel assembly is dismantled and a portion of withdrawn fuel rod is air-tightly sealed with an inspection portion. The inside of the inspection portion is maintained at a pressure-reduced state. Then, in a case if failed openings are formed at a portion sealed by the inspection portion in the fuel rod, FP gases in the fuel rod are released based on the reduced pressure and the FP gases are detected in the detection portion. Accordingly, by relatively moving the detection portion to the fuel rod, the failure position can be detected. (Yoshino, Y.).

Crud deposition modeling on BWR fuel rods

International Nuclear Information System (INIS)

Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

2014-01-01

Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

Tools for LWR spent fuel characterization: Assembly classes and fuel designs

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1991-01-01

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

Automated Fuel Element Closure Welding System

International Nuclear Information System (INIS)

Wahlquist, D.R.

1993-01-01

The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

26 CFR 1.475(b)-2 - Exemptions-identification requirements.

Science.gov (United States)

2010-04-01

... requirements. (a) Identification of the basis for exemption. An identification of a security as exempt from mark to market does not satisfy section 475(b)(2) if it fails to state whether the security is described in— (1) Either of the first two subparagraphs of section 475(b)(1) (identifying a security as held...

Fuel failure assessments based on radiochemistry. Experience feedback and challenges

International Nuclear Information System (INIS)

Petit, C.; Ziabletsev, D.; Zeh, P.

2015-01-01

Significant improvements have been observed in LWR nuclear fuel reliability over the past years. As a result, the number of fuel failures in PWRs and BWRs has recently dramatically decreased. Nevertheless, a few remaining challenges still exist. One of them is that the industry has recently started seeing a relatively new type of fuel failure, so-called 'weak leak failures', which could be characterized by a very small release of gaseous fission products and essentially almost zero release of iodines or any other soluble fission products in the reactor coolant. Correspondingly, the behavior of these weak leakers does not follow typical behavior of a conventional leaker characterized by a proportionality of the amount of released Xe 133 related to the failed rod power. Instead, for a weak leaker, the activity of Xe 133 is directly correlated to the size of the cladding defects of the leaker. The presence of undetected weak leaker in the core may lead to carryover of a leaker into the subsequent cycle. Even if the presence of weak leaker in the core is suspected, it typically requires more effort to identify the leaker which could result in extended duration of the outage and ultimately to economic losses to the utility operating the reactor. To effectively deal with this issue the industry has been facing, several changes have been recently realized, which are different from the methodology of dealing with conventional leaker. These changes include new assessment methods, the need for improved sipping techniques to better identify low release leakers, and correspondingly better equipment to be able to locate small clad defects associated with weak leaker, such as sensitive localization device of failed rods, sensitive eddy current coil for the failed rod, ultra high definition cameras for the failed rod examination and experienced fuel reliability engineers performing cause of failure and rood cause research and analyses. Ultimately, the destructive

High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-12-01

High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

The characterization and monitoring of metallic fuel breaches in EBR-2

International Nuclear Information System (INIS)

Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

1991-01-01

This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

Nuclear fuel assembly

International Nuclear Information System (INIS)

Takeda, Tadashi; Sato, Kenji; Goto, Masakazu.

1984-01-01

Purpose: To facilitate identification of a fuel assembly upon fuel exchange in BWR type reactors. Constitution: Fluorescent material is coated or metal plating is applied to the impressed portion of a upper tie plate handle of a fuel assembly, and the fluorescent material or the metal plating surface is covered with a protective membrane made of transparent material. This enables to distinguish the impressed surface from a distant place and chemical reaction between the impressed surface and the reactor water can be prevented. Furthermore, since the protective membrane is formed such that it protrudes toward the upper side relative to the impressed surface, there is no risk of depositions of claddings thereover. (Moriyama, K.)

State of the art: Multi-fuel reformers for automotive fuel cell applications. Problem identification and research needs

Energy Technology Data Exchange (ETDEWEB)

Westerholm, R. [Stockholm Univ. (Sweden). Dept. of Analytical Chemistry; Pettersson, L.J. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemical Engineering and Technology

1999-12-01

On an assignment from the Transport and Communications Research Board (KFB) a literature study and a study trip to the USA and Great Britain have been performed. The literature study and the study trip was made during late spring and autumn 1999.The purpose of the project was to collect available information about the chemical composition of the product gas from a multi-fuel reformer for a fuel cell vehicle. It was furthermore to identify problems and research needs. The report recommends directions for future major research efforts. The results of the literature study and the study trip led to the following general conclusions: With the technology available today it does not seem feasible to develop a highly efficient and reliable multi-fuel reformer for automotive applications, i. e. for applications where all types of fuels ranging from natural gas to heavy diesel fuels can be used. The potential for developing a durable and reliable system is considerably higher if dedicated fuel reformers are used.The authors propose that petroleum-derived fuels should be designed for potential use in mobile fuel cell applications. In the present literature survey and the site visit discussions we found that there are relatively low emissions from fuel cell engines compared to internal combustion engines. However, the major research work on reformers/fuel cells have been performed during steady-state operation. Emissions during start-up, shutdown and transient operation are basically unknown and must be investigated in more detail. The conclusions and findings in this report are based on open/available information, such as discussions at site visits, reports, scientific publications and symposium proceedings.

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

International Nuclear Information System (INIS)

Marshall, William J.; Wagner, John C.

2012-01-01

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., 45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k eff . Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

Test Plan for Cask Identification Detector

Energy Technology Data Exchange (ETDEWEB)

Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-29

This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

Inserts for nuclear fuel elements

International Nuclear Information System (INIS)

Cragg, P.J.

1982-01-01

An insert for a nuclear fuel pin which comprises a strip. The strip carries notches, which enable a coding arrangement to be carried on the strip. The notches may be of differing sizes and the coding on the strip includes identification and identification checking data. Each notch on the strip may give rise to a signal pulse which is counted by a detector to avoid errors. (author)

Method of locating a leaking fuel element in a fast breeder power reactor

International Nuclear Information System (INIS)

Honekamp, J.R.; Fryer, R.M.

1976-01-01

Failed fuel detection and location in an FBR is accomplished by mass spectrometric analysis of the cover gas for the 134 Xe/ 133 Xe ratio, and correlation with the theoretical ratio in fuel elements of known power level and burnup. (E.C.B.)

Fuel spill identification by gas chromatography -- genetic algorithms/pattern recognition techniques

International Nuclear Information System (INIS)

Lavine, B.K.; Moores, A.J.; Faruque, A.

1998-01-01

Gas chromatography and pattern recognition methods were used to develop a potential method for typing jet fuels so a spill sample in the environment can be traced to its source. The test data consisted of 256 gas chromatograms of neat jet fuels. 31 fuels that have undergone weathering in a subsurface environment were correctly identified by type using discriminants developed from the gas chromatograms of the neat jet fuels. Coalescing poorly resolved peaks, which occurred during preprocessing, diminished the resolution and hence information content of the GC profiles. Nevertheless a genetic algorithm was able to extract enough information from these profiles to correctly classify the chromatograms of weathered fuels. This suggests that cheaper and simpler GC instruments ca be used to type jet fuels

Off-gas monitor system for the detection of a failed fuel cladding tube

International Nuclear Information System (INIS)

Yuasa, Yoshiyuki; Oosaki, Masahiko; Naito, Makoto.

1984-01-01

Purpose: To rapidly and reliably detect failures in a fuel cladding tube thereby prevent accidents. Method: Off-gases discharged from an air extractor are mixed at a certain ratio with cleaning water and gamma spectra for short-life radioactive rare gases and for long-life radioactive rare gases in the off-gases are measured by gamma ray detectors. The spectra are analyzed by a Pulse-height analyzer so as to quantitatively determine the nuclides of radioactive rare gases thereby calculate the release rate on each of the nuclides. Further, a central data-processing unit calculates the composition ratio of each release models about recoil, diffusion and equilibrium in the whole radioactive rare gases based on the release rate. As the total value of the release rate is rapidly increased, recoil part will be decreased suddenly and the diffusion and equilibrium parts will be increased by so much, upon failure of a fuel can; thus the failure can rapidly be detected. (Sekiya, K.)

Engaging Future Failing States

Science.gov (United States)

2011-03-23

military missions in the Middle East, the Balkans, Africa, Asia , and South America. There is an increasing proliferation of failed and failing states...disparity, overpopulation , food security, health services availability, migration pressures, environmental degradation, personal and 22 community

Posttest examination results of recent treat tests on metal fuel

International Nuclear Information System (INIS)

Holland, J.W.; Wright, A.E.; Bauer, T.H.; Goldman, A.J.; Klickman, A.E.; Sevy, R.H.

1986-01-01

A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins

Fuel exchange device for FBR type reactor

International Nuclear Information System (INIS)

Onuki, Koji.

1993-01-01

The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

Method of detecting fuel failure in FBR type reactor and method of estimating fuel failure position

International Nuclear Information System (INIS)

Sonoda, Yukio; Tamaoki, Tetsuo

1989-01-01

Noise components in a normal state contained in detection signals from delayed neutron monitors disposed to a coolant inlet, etc. of an intermediate heat exchanger are forecast by self-recurring model and eliminated, and resultant detection signals are monitored thereby detecting fuel failure high sensitivity. Subsequently, the reactor is controlled to a low power operation state and a new self-recurring model to the detection signals from the delayed neutron monitors are prepared. Then, noise components in this state are removed and control rods near the delayed neutron monitors are extracted in a short stroke successively to examine the change of response of the delayed neutron monitors. Accordingly, the failed position for each of the fuels can be estimated at a level of one fuel assembly or a level of several assemblies containing the above-mentioned fuel assembly. Since the fuel failure can be detected at a high sensitivity and the position can be estimated, diffusion of abnormality can be prevented and plant shutdown for fuel exchange can be minimized. (I.S.)

The role of fission products (noble metal particles) in spent fuel corrosion process in a failed container

Energy Technology Data Exchange (ETDEWEB)

Wu, L., E-mail: lwu59@uwo.ca [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada); Univ. of Western Ontario, Surface Science Western, London, Ontario (Canada)

2013-07-01

The corrosion/dissolution of simulated spent fuel has been studied electrochemically. Fission products within the UO{sub 2} matrix are found to have significant effect on the anodic dissolution behaviour of the fuel. It is observed that H{sub 2}O{sub 2}oxidation is accelerated on the surfaces of doped noble metal (ε) particles existing in the fuel matrix. It is concluded that H{sub 2}O{sub 2} decomposition rather than UO{sub 2} corrosion should be the dominant reaction under high H{sub 2}O{sub 2} concentrations. (author)

Fuel Behavior Modeling Issues Associated with Future Fast Reactor Systems

International Nuclear Information System (INIS)

Yacout, A.M.; Hofman, G.L.; Lambert, J.D.B.; Kim, Y.S.

2007-01-01

Major issues of concern related to advanced fast reactor fuel behavior are discussed here with focus on phenomena that are encountered during irradiation of metallic fuel elements. Identification of those issues is part of an advanced fuel simulation effort that aims at improving fuel design and reducing reliance on conventional approach of design by experiment which is both time and resource consuming. (authors)

Fuel canister and blockage pin fabrication for SLSF Experiment P4

International Nuclear Information System (INIS)

Rhude, H.V.; Folkrod, J.R.; Noland, R.A.; Schaus, P.S.; Benecke, M.W.; Delucchi, T.A.

1983-01-01

As part of its fast breeder reactor safety research program, Argonne National Laboratory (ANL) has conducted an experiment (SLSF Experiment P4) to determine the extent of fuel-failure propagation resulting from the release of molten fuel from one or more heat-generating fuel canisters. The test conditions consisted of 37 full-length FTR fuel pins operating at FTR rated core nominal peak fuel/reduced coolant conditions. Thirty-four of the the fuel pins were prototypical FTR mixed-oxide fuel pins. The other three fuel pins were fabricated with a mid-core section having an enlarged canister containing fully enriched UO 2 . Two of the canisters were cylindrical and one was fluted. The cylindrical canisters were designed to fail and release molten fuel into the 37-pin fuel cluster at near full power

Conditions inside Water Pooled in a Failed Nuclear Waste Container and its Effect on Radionuclide Release

Science.gov (United States)

Hamdan, L. K.; Walton, J. C.; Woocay, A.

2009-12-01

Nuclear power use is expected to expand in the future, as part of the global clean energy initiative, to meet the world’s surging energy demand, and attenuate greenhouse gas emissions, which are mainly caused by fossil fuels. As a result, it is estimated that hundreds of thousands of metric tons of spent nuclear fuel (SNF) will accumulate. SNF disposal has major environmental (radiation exposure) and security (nuclear proliferation) concerns. Storage in unsaturated zone geological repositories is a reasonable solution for dealing with SNF. One of the key factors that determine the performance of the geological repository is the release of radionuclides from the engineered barrier system. Over time, the nuclear waste containers are expected to fail gradually due to general and localized corrosions and eventually infiltrating water will have access to the nuclear waste. Once radionuclides are released, they will be transported by water, and make their way to the accessible environment. Physical and chemical disturbances in the environment over the container will lead to different corrosion rates, causing different times and locations of penetration. One possible scenario for waste packages failure is the bathtub model, where penetrations occur on the top of the waste package and water pools inside it. In this paper the bathtub-type failed waste container is considered. We shed some light on chemical and physical processes that take place in the pooled water inside a partially failed waste container (bathtub category), and the effects of these processes on radionuclide release. Our study considers two possibilities: temperature stratification of the pooled water versus mixing process. Our calculations show that temperature stratification of the pooled water is expected when the waste package is half (or less) filled with water. On the other hand, when the waste package is fully filled (or above half) there will be mixing in the upper part of water. The effect of

Method to mount defect fuel elements i transport casks

International Nuclear Information System (INIS)

Borgers, H.; Deleryd, R.

1996-01-01

Leaching or otherwise failed fuel elements are mounted in special containers that fit into specially designed chambers in a transportation cask for transport to reprocessing or long-time storage. The fuel elements are entered into the container under water in a pool. The interior of the container is dried before transfer to the cask. Before closing the cask, its interior, and the exterior of the container are dried. 2 figs

Fuel starvation. Irreversible degradation mechanisms in PEM fuel cells

Energy Technology Data Exchange (ETDEWEB)

Rangel, Carmen M.; Silva, R.A.; Travassos, M.A.; Paiva, T.I.; Fernandes, V.R. [LNEG, National Laboratory for Energy and Geology, Lisboa (Portugal). UPCH Fuel Cells and Hydrogen Unit

2010-07-01

PEM fuel cell operates under very aggressive conditions in both anode and cathode. Failure modes and mechanism in PEM fuel cells include those related to thermal, chemical or mechanical issues that may constrain stability, power and lifetime. In this work, the case of fuel starvation is examined. The anode potential may rise to levels compatible with the oxidization of water. If water is not available, oxidation of the carbon support will accelerate catalyst sintering. Diagnostics methods used for in-situ and ex-situ analysis of PEM fuel cells are selected in order to better categorize irreversible changes of the cell. Electrochemical Impedance Spectroscopy (EIS) is found instrumental in the identification of fuel cell flooding conditions and membrane dehydration associated to mass transport limitations / reactant starvation and protonic conductivity decrease, respectively. Furthermore, it indicates that water electrolysis might happen at the anode. Cross sections of the membrane catalyst and gas diffusion layers examined by scanning electron microscopy indicate electrode thickness reduction as a result of reactions taking place during hydrogen starvation. Catalyst particles are found to migrate outwards and located on carbon backings. Membrane degradation in fuel cell environment is analyzed in terms of the mechanism for fluoride release which is considered an early predictor of membrane degradation. (orig.)

Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

International Nuclear Information System (INIS)

Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin

1996-07-01

Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author)

IFR fuel cycle--pyroprocess development

International Nuclear Information System (INIS)

Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

1992-01-01

The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

Preliminary concepts: safeguards for spent light-water reactor fuels

International Nuclear Information System (INIS)

Cobb, D.D.; Dayem, H.A.; Dietz, R.J.

1979-06-01

The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems

System of leak inspection of irradiated fuel

International Nuclear Information System (INIS)

Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F.

2007-01-01

The International Atomic Energy Agency (IAEA) through the project RLA/04/18 Irradiated Fuel Management in Research reactors, recommended among other that the participant countries (Brazil, Argentina, Chile, Peru and Mexico), develop the sipping tool to generate registrations of the state that keep the irradiated fuels in the facilities of each country. The TRIGA Mark lll Reactor (RTMIII) Department, generated a project that it is based on the dimensions of the used fuel by the RTMIII, for design and to build an inspection system of irradiated fuel well known as SIPPING. This technique, provides a high grade of accuracy in the detection of gassy fission products or liquids that escape from the enveloping of fuels that have flaws or flights. The operation process of the SIPPING is carried out generating the migration of fission products through the creation of a pressure differential gas or vacuum to identify fuel assemblies failed by means of the detection of the xenon and/or krypton presence. The SIPPING system, is a device in revolver form with 4 tangential nozzles, which will discharge the fluid between the external surface of the enveloping of the fuel and the interior surface of the encircling one; the device was designed with independent pieces, with threaded joining and with stamps to impede flights of the fluid toward the exterior of the system. The System homogenizes and it distributes the fluid pressure so that the 4 nozzles work to equality of conditions, for what the device was designed in 3 pieces, an internal that is denominated revolver, one external that calls cover, and a joining called mamelon that will unite with the main encircling of the system. The detection of fission products in failed fuels, its require that inside the encircling one where the irradiated fuel element is introduced, be generated a pressure differential of gas or vacuum, and that it allows the samples extraction of water. For what generated a top for the encircling with the

Fast breeder fuel element development

International Nuclear Information System (INIS)

Marth, W.; Muehling, G.

1983-08-01

This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

24 CFR 5.218 - Penalties for failing to disclose and verify Social Security and Employer Identification Numbers.

Science.gov (United States)

2010-04-01

... and verify Social Security and Employer Identification Numbers. 5.218 Section 5.218 Housing and Urban... REQUIREMENTS; WAIVERS Disclosure and Verification of Social Security Numbers and Employer Identification Numbers; Procedures for Obtaining Income Information Disclosure and Verification of Social Security...

Stress and Diffusion in Stored Pu ZPPR Fuel from Alpha Generation

Energy Technology Data Exchange (ETDEWEB)

Charles W. Solbrig; Chad L. Pope; Jason P. Andrus

2014-07-01

ZPPR (Zero Power Physics Reactor) is a research reactor that has been used to investigate breeder reactor fuel designs. The reactor has been dismantled but its fuel is still stored there. Of concern are its plutonium containing metal fuel elements which are enclosed in stainless steel cladding with gas space filled with helium–argon gas and welded air tight. The fuel elements which are 5.08 cm by 0.508 cm up to 20.32 cm long (2 in × 0.2 in × 8 in) were manufactured in 1968. A few of these fuel elements have failed releasing contamination raising concern about the general state of the large number of other fuel elements. Inspection of the large number of fuel elements could lead to contamination release so analytical studies have been conducted to estimate the probability of failed fuel elements. This paper investigates the possible fuel failures due to generation of helium in the metal fuel from the decay of Pu and its possible damage to the fuel cladding from metal fuel expansion or from diffusion of helium into the fuel gas space. This paper (1) calculates the initial gas loading in a fuel element and its internal free volume after it has been brought into the atmosphere at ZPPR, (2) shows that the amount of helium generated by decay of Pu over 46 years since manufacture is significantly greater than this initial loading, (3) determines the amount of fuel swelling if the helium stays fixed in the fuel plate and estimates the amount of helium which diffuses out of the fuel plate into the fuel plenum assuming the helium does not remain fixed in the fuel plate but can diffuse to the plenum and possibly through the cladding. Since the literature is not clear as to which possibility occurs, as with Schroedinger’s cat, both possibilities are analyzed. The paper concludes that (1) if the gas generated is fixed in the fuel, then the fuel swelling it can cause would not cause any fuel failure and (2) if the helium does diffuse out of the fuel (in accordance

Development of a Computer Code for the Estimation of Fuel Rod Failure

Energy Technology Data Exchange (ETDEWEB)

Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

1997-12-31

Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

FY2015 ceramic fuels development annual highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

2015-09-22

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

FY2016 Ceramic Fuels Development Annual Highlights

Energy Technology Data Exchange (ETDEWEB)

Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-01-24

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

International Nuclear Information System (INIS)

Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-01-01

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

2015-04-15

Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

Dams designed to fail

Energy Technology Data Exchange (ETDEWEB)

Penman, A. [Geotechnical Engineering Consultants, Harpenden (United Kingdom)

2004-09-01

New developments in geotechnical engineering have led to methods for designing and constructing safe embankment dams. Failed dams can be categorized as those designed to fail, and those that have failed unexpectedly. This presentation outlined 3 dam failures: the 61 m high Malpasset Dam in France in 1959 which killed 421; the 71 m high Baldwin Hills Dam in the United States in 1963 which killed 5; and, the Vajont Dam in Italy in 1963 which killed 2,600 people. Following these incidents, the International Commission for Large Dams (ICOLD) reviewed regulations on reservoir safety. The 3 dams were found to have inadequate spillways and their failures were due to faults in their design. Fuse plug spillways, which address this problem, are designed to fail if an existing spillway proves inadequate. They allow additional discharge to prevent overtopping of the embankment dam. This solution can only be used if there is an adjacent valley to take the additional discharge. Examples of fuse gates were presented along with their effect on dam safety. A research program is currently underway in Norway in which high embankment dams are being studied for overtopping failure and failure due to internal erosion. Internal erosion has been the main reason why dams have failed unexpectedly. To prevent failures, designers suggested the use of a clay blanket placed under the upstream shoulder. However, for dams with soft clay cores, these underblankets could provide a route for a slip surface and that could lead to failure of the upstream shoulder. It was concluded that a safe arrangement for embankment dams includes the use of tipping gates or overturning gates which always fail at a required flood water level. Many have been installed in old and new dams around the world. 14 refs., 19 figs.

Predicting failing performance on a standardized patient clinical performance examination: the importance of communication and professionalism skills deficits.

Science.gov (United States)

Chang, Anna; Boscardin, Christy; Chou, Calvin L; Loeser, Helen; Hauer, Karen E

2009-10-01

The purpose is to determine which assessment measures identify medical students at risk of failing a clinical performance examination (CPX). Retrospective case-control, multiyear design, contingency table analysis, n = 149. We identified two predictors of CPX failure in patient-physician interaction skills: low clerkship ratings (odds ratio 1.79, P = .008) and student progress review for communication or professionalism concerns (odds ratio 2.64, P = .002). No assessments predicted CPX failure in clinical skills. Performance concerns in communication and professionalism identify students at risk of failing the patient-physician interaction portion of a CPX. This correlation suggests that both faculty and standardized patients can detect noncognitive traits predictive of failing performance. Early identification of these students may allow for development of a structured supplemental curriculum with increased opportunities for practice and feedback. The lack of predictors in the clinical skills portion suggests limited faculty observation or feedback.

Fatty Acid Oxidation Is Preserved Regardless of Impaired Uptake in the Chronically Failing Rat Heart

OpenAIRE

TACHIKAWA, Hitoshi

2004-01-01

Fatty acid is used as a major fuel in the fasting heart, but the precise metabolism in the failing heart remains unknown. We assessed the hypothesis that the fatty acid metabolism might be impaired or delayed during heart failure. We examined in vivo kinetics of an isotope-labeled fatty acid analogue and its substrates as well as hemodynamic parameters and histopathological findings in a rat model of postmyocarditic dilated cardiomyopathy. Rat experimental autoimmune myocarditis (EAM) was ind...

Failed endotracheal intubation

Directory of Open Access Journals (Sweden)

Sheykhol Islami V

1995-07-01

Full Text Available The incidence of failed intubation is higher in obstetric than other surgical patients. Failed intubation was the 2nd commonest cause of mortality during anesthesia. Bearing in mind that failre to intubate may be unavoidable in certain circumstances, it is worth reviewing. The factors, which may contribute to a disastrous out come. Priorities of subsequent management must include maintaining oxygenation and preventing aspiration of gastric contents. Fiber optic intubation is now the technique of choice with a high success rate and with least trauma to the patient.

Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

International Nuclear Information System (INIS)

Misfeldt, I.

1980-01-01

A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

Development of evaluation method of fuel failure fraction during the High Temperature Engineering Test Reactor operation

Energy Technology Data Exchange (ETDEWEB)

Sawa, Kazuhiro; Yoshimuta, Shigeharu; Tobita, Tsutomu; Sato, Masashi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1997-05-01

The High Temperature Engineering Test Reactor (HTTR) uses coated particles as fuel. During normal operation, short-lived noble gases are mainly released by diffusion from fuel particles with defects in their coating layers (i.e., failed particle). Since noble gases do not plate out on the inner surfaces of primary cooling system, their activities in primary coolant reflect fuel failure fraction in the core. An evaluation method was developed to predict failure fraction of coated fuel particles during normal operation of the HTTR. The method predicts core-average and hot plenum regionwise failure fractions based on the fractional releases, (R/B)s, of noble gases. The (R/B)s are calculated by fission gas concentration measurements in the primary cooling system of the HTTR. Recent fabrication data show that through-coatings failure fraction is extremely low. Then, fractional release from matrix contamination uranium, which is background for accurate evaluation of the fuel failure fraction, should be precisely predicted. This report describes an evaluation method of fuel failure fraction from measurements in the HTTR together with a fission gas release model from fuel compact containing failed particles and matrix contamination uranium. (author)

Evaluation of the effect of probe design parameters on ECT signal and development of eddy current probe for irradiated fuel rods

International Nuclear Information System (INIS)

Kwank, S. W.; Han, Y. K.; Woo, S. K.; Kim, T. W.; Park, J. Y.; Kim, B. J.; Park, J. Y.

1999-01-01

Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

Fuel Retrieval and Management of Fuel Element Debris

International Nuclear Information System (INIS)

Chande, Shridhar; Lachaume, J. L.

2013-01-01

Nuclear accidents involving core meltdown have not been so rare. While the first occurred in early fifties, it is reported that about 20 have occurred worldwide in military and commercial reactors. The more recent and major accidents are 1. Three Mile Island, USA in 1979: Approximately half the core was melted, and flowed to the bottom of the reactor pressure vessel however the pressure vessel remained intact and contained the damaged fuel. 2. Chernobyl, former USSR in 1984: Explosive release of radioactive material occurred. About 6 tons of fuel was dispersed as air-borne particles. Most of the core was damaged or melted. 3. Fukushima, Japan 2011: Three units suffered melt down. In unit 1 almost all the fuel assemblies melted and accumulated at the bottom of the vessel. It is reported that the vessel failed and the molten corium has penetrated the concrete. In the units 2 and 3, partial melting of cores has occurred. In several of these cases, fuel retrieval and management activities have been carried out. The experience and insights gained from these activities will be extremely useful for planning and execution of similar activities in future if ever they are needed. The purpose of this session was to exchange this experience and also to share the lessons learned. This is of particularly important, at this juncture, when planning and preparation for retrieval of damaged cores in Fukushima NPP is in progress. (author)

Handling apparatus for fuel assemblies in a core

International Nuclear Information System (INIS)

Hatakenaka, Hideo.

1975-01-01

Object: To prevent an occurrence of a cloud as well as trouble in outflow of cooling water at the time of failure, in a window through which the operation of a collet installing and removing mechanism is monitored. Structure: A monitoring window comprises a pair of transparent window panes between which is interposed a non-compressive transparent fluid. With this construction, when the collet installing and removing mechanism within a container is operated while illuminating it by light means and monitoring it by a television camera to connect a fuel assembly with a shielding plug, and even if one transparent window pane should be failed as a result of trouble, the other transparent window pane prevents outflow of cooling water within a fuel transferring transfer port, and at the same time, the scattering force of fragments of failed transparent window pane is attenuated by the non-compressive transparent body within the monitoring window chamber. (Hanada, M.)

Interim report spent nuclear fuel retrieval system fuel handling development testing

Energy Technology Data Exchange (ETDEWEB)

Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

1997-06-01

Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

International Nuclear Information System (INIS)

Peacock, H.B.

1983-08-01

Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 8 -aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 0 C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

Fuel poverty is facing a data challenge - Which strategies to struggle against fuel poverty? Propositions for an ecologic and social transition policy

International Nuclear Information System (INIS)

Erard, Timothee; Chancel, Lucas; Saujot, Mathieu

2015-01-01

The authors address three main questions: how are structured the network of actors and the tools for the struggle against fuel poverty, what are the specific data challenges faced by the actors and the possible responses, and which lessons can be learned for the governance of energy transition in its whole. After a presentation of the context of fuel poverty (analysis of tools for the struggle against fuel poverty, the use of social-energetic data), this study, based on about forty interviews of various actors and on a workshop, proposes an analysis framework which distinguishes six steps in the definition and implementation of policies of struggle against fuel poverty. After a description of the current status and an identification of required improvements for each step, the authors propose a set of recommendations, draw lessons for urban policies aimed at an ecologic transformation and a modernisation of the social protection system in terms of level of intervention, scope of actions, and ownership and access to data bases. These recommendations more particularly address the definition of fuel poverty, its diagnosis at the national and at the territorial level, a better identification of concerned households, and an assessment of existing arrangements

Operational reliability testing of FBR fuel in EBR-II

International Nuclear Information System (INIS)

Asaga, Takeo; Ukai, Shigeharu; Nomura, Shigeo; Shikakura, Sakae

1991-01-01

The operational reliability testing of FBR fuel has been conducting in EBR-II as a DOE/PNC collaboration program. This paper reviews the achieved summary of Phase-I test as well as outline of progressing Phase-II test. In Phase-I test, the reliability of FBR fuel pins including 'MONJU' fuel was demonstrated at the event of operational transient. Continued operation of the failed pins was also shown to be feasible without affecting the plant operation. The objectives of the Phase-II test is to extend the data base relating with the operational reliability for long life fuel, and to supply the highly quantitative evaluation. The valuable insight obtained in Phase-II test are considerably expected to be useful toward the achievement of commercial FBR. (author)

Experiments to investigate the effects of small changes in fuel stoichiometry on fission gas release

Energy Technology Data Exchange (ETDEWEB)

Copeland, P S; Smith, R C [Windscale Lab., AEA Technology, Seascale, Cumbria (United Kingdom)

1997-08-01

Fuel pin failure in-reactor leads to fission product and in the case of a PWR fuel debris release to the coolant. For economic reasons immediate shutdown and discharge of failed fuel needs to be avoided but this needs to be counter-balanced against the increasing dose to operators. PWR practice is to continue running wit failed rods, monitoring coolant activity, and only shutting down the reactor and discharging the fuel when circuit activity levels become unacceptable. The rate of fission product release under failed fuel conditions is of key importance and considerable effort has been directed towards establishing the dependency of release on temperature, heating rate, burn-up, and also the extent of fuel oxidation. As a precursor to a possible wider investigation of this area, a small programme was mounted during 1992/1993 to confirm whether small changes in the oxidation state of the fuel, for example those caused by minor cladding defects, would significantly effect fuel behaviour during postulated design basis faults. The objective of the programme was to determine the effects of small departures from stoichiometric fuel composition on fission gas release, and to compare the results with the current methodology for calculating releases under fault conditions. A total of eight experiments was performed. Two were intended as baseline tests to provide a reference with which to compare the effect of oxidation state influenced behaviour with that of thermal effects. It was found that small changes in stoichiometry of {sup {approx}}1 x 10{sup -6} had little or no effect on release but that changes of {sup {approx}} 1 x 10{sup -4} were observed to increase the diffusion coefficient, for {sup 85}Kr, by up to an order of magnitude and hence greatly increase the release rate. The stoichiometry of the sample used in these tests was, for convenience, adjusted using He/H{sub 2}/H{sub 2}O atmospheres. (Abstract Truncated)

European experience in the transport of irradiated light-water reactor fuel

International Nuclear Information System (INIS)

Curtis, H.W.

1979-01-01

Various methods of transport of irradiated fuel flasks in Europe are described. While many problems in the transport of heavy flasks have been solved some remain and new ones have appeared. Some of these problems are the accumulation of crud on the surface of fuel elements, the problems of failed fuel, stringent criticality criteria, the ''sweating out'' of contaminated flasks, the access, road or rail, to reactor sites, and the maintenance of the transport vehicles. Some future trends in the direction of heavy flasks in the range of 75 to 100 tonnes are indicated

Performance assessment analyses unique to Department of Energy spent nuclear fuel

International Nuclear Information System (INIS)

Loo, H.H.; Duguid, J.J.

2000-01-01

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

Investigation of the ramp testing behaviour of fuel pins with different diameters

International Nuclear Information System (INIS)

Pott, G.; Herren, M.; Wigger, B.

1979-09-01

The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

Fact sheet on fuel manufacturing and utilization

International Nuclear Information System (INIS)

2006-01-01

The Nuclear Fuel Cycle and Materials Section (NFCMS) supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle, provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection, supports the development fuel modeling expertise in Member States, covering both normal operation and postulated and severe accident conditions, provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation, supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, provides information and support research into the basic properties of fuel materials, including UO2, MOX, (Th, Pu)O2, (Th, U233)O2 fuels and zirconium alloy cladding and fuel assembly components and offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology and materials, economic and other aspects of nuclear fuel use (e.g. environmental impact). Recently NFCMS provided support to a Member State manufacturing Gadolinia doped fuel and provided in-mast sipping equipment to a Nuclear Power Plant to allow the determination of fuel failure. Member States interested in fuel performance and manufacture should contact the Technical Cooperation Department of the Agency and Member States interested in knowing more about the Agency's programme on source management should contact: C. Ganguly, Section Head, V. Inozemtsev, J. Killeen

Exploration for fossil and nuclear fuels from orbital altitudes

Science.gov (United States)

Short, N. M.

1977-01-01

The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

Critical thinking: are the ideals of OBE failing us or are we failing the ...

African Journals Online (AJOL)

Critical thinking: are the ideals of OBE failing us or are we failing the ideals of OBE? K Lombard, M Grosser. Abstract. No Abstract. South African Journal of Education Vol. 28 (4) 2008: pp. 561-580. Full Text: EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT · DOWNLOAD FULL TEXT DOWNLOAD FULL TEXT.

Fusion fuel cycle solid radioactive wastes

International Nuclear Information System (INIS)

Gore, B.F.; Kaser, J.D.; Kabele, T.J.

1978-06-01

Eight conceptual deuterium-tritium fueled fusion power plant designs have been analyzed to identify waste sources, materials and quantities. All plant designs include the entire D-T fuel cycle within each plant. Wastes identified include radiation-damaged structural, moderating, and fertile materials; getter materials for removing corrosion products and other impurities from coolants; absorbents for removing tritium from ventilation air; getter materials for tritium recovery from fertile materials; vacuum pump oil and mercury sludge; failed equipment; decontamination wastes; and laundry waste. Radioactivity in these materials results primarily from neutron activation and from tritium contamination. For the designs analyzed annual radwaste volume was estimated to be 150 to 600 m 3 /GWe. This may be compared to 500 to 1300 m 3 /GWe estimated for the LMFBR fuel cycle. Major waste sources are replaced reactor structures and decontamination waste

Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

International Nuclear Information System (INIS)

Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

1979-01-01

Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

Molybdenum-base cermet fuel development

International Nuclear Information System (INIS)

Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

1987-07-01

Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

Numerical analyses of an ex-core fuel incident: Results of the OECD-IAEA Paks Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hozer, Z., E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Aszodi, A. [BME NTI Budapest (Hungary); Barnak, M. [IVS, Trnava (Slovakia); Boros, I. [BME NTI Budapest (Hungary); Fogel, M. [VUJE, Trnava (Slovakia); Guillard, V. [IRSN, Cadarache (France); Gyori, Cs. [ITU, EU, Karlsruhe (Germany); Hegyi, G. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Horvath, G.L. [VEIKI, Budapest (Hungary); Nagy, I. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Junninen, P. [VTT, Espoo (Finland); Kobzar, V. [KI, Moscow (Russian Federation); Legradi, G. [BME NTI Budapest (Hungary); Molnar, A. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Pietarinen, K. [VTT, Espoo (Finland); Perneczky, L. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Makihara, Y. [ATMEA, Paris (France); Matejovic, P. [IVS, Trnava (Slovakia); Perez-Fero, E.; Slonszki, E. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary)

2010-03-15

The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data. The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 deg. C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg.

Evaluation of fuel release rate and mechanism tests under RBCB conditions

International Nuclear Information System (INIS)

Adamson, M.G.

1981-09-01

This task includes theoretical evaluation of fuel/fission product release behavior from failed LMFBR fuel elements as well as an on-going experimental investigation of the mechanism of oxide fuel dispersal into flowing liquid sodium. The primary objectives of this work are to develop a fuel source term that can be used in predictive models for primary heat transfer system contamination and to understand the separate influences of important system variables (such as flow rate, oxygen impurity level) on this source term. The present report is written in two parts: the first, in condensed form, is an updated evaluation of fuel (U,Pu) and fission product release data, and the second describes the current status of supporting experimental work at General Electric's Vallecitos Laboratory

Tennessee Valley Authority nuclear fuel performance

International Nuclear Information System (INIS)

Rahn, B.; Gardner, D.

1985-01-01

The Browns Ferry BWR units started operation in the early to mid 1970s, and the Sequoyah PWR units started in the very early 1980s. Operating dates, outage length, capacity factor, and cycle exposure are given for these units. This paper deals with fuel performance at the Browns Ferry and Sequoyah plants. Leaks, off-gas, fraction failed, iodine-131, and corrosion are covered. 14 figs

Light-duty vehicle fuel economy improvements, 1979--1998: A consumer purchase model of corporate average fuel economy, fuel price, and income effects

Science.gov (United States)

Chien, David Michael

2000-10-01

The Energy Policy and Conservation Act of 1975, which created fuel economy standards for automobiles and light trucks, was passed by Congress in response to the rapid rise in world oil prices as a result of the 1973 oil crisis. The standards were first implemented in 1978 for automobiles and 1979 for light trucks, and began with initial standards of 18 MPG for automobiles and 17.2 MPG for light trucks. The current fuel economy standards for 1998 have been held constant at 27.5 MPG for automobiles and 20.5 MPG for light trucks since 1990--1991. While actual new automobile fuel economy has almost doubled from 14 MPG in 1974 to 27.2 MPG in 1994, it is reasonable to ask if the CAFE standards are still needed. Each year Congress attempts to pass another increase in the Corporate Average Fuel Economy (CAFE) standard and fails. Many have called for the abolition of CAFE standards citing the ineffectiveness of the standards in the past. In order to determine whether CAFE standards should be increased, held constant, or repealed, an evaluation of the effectiveness of the CAFE standards to date must be established. Because fuel prices were rising concurrently with the CAFE standards, many authors have attributed the rapid rise in new car fuel economy solely to fuel prices. The purpose of this dissertation is to re-examine the determinants of new car fuel economy via three effects: CAFE regulations, fuel price, and income effects. By measuring the marginal effects of the three fuel economy determinants upon consumers and manufacturers choices, for fuel economy, an estimate was made of the influence of each upon new fuel economy. The conclusions of this dissertation present some clear signals to policymakers: CAFE standards have been very effective in increasing fuel economy from 1979 to 1998. Furthermore, they have been the main cause of fuel economy improvement, with income being a much smaller component. Furthermore, this dissertation has suggested that fuel prices have

Assessment of Research Needs for Advanced Fuel Cells

Energy Technology Data Exchange (ETDEWEB)

Penner, S.S.

1985-11-01

The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

Experimental and inspection facilities in post-irradiation of spent fuel pools for the analysis of the behaviour of nuclear fuels in power reactors

International Nuclear Information System (INIS)

Ruggirello, G.; Zawerucha, A.

1992-01-01

Since the beginning of the Atomic Nuclear Reactors (PHWR) Atucha I and Embalse in Argentine are employed different techniques for the knowing of the fuel bundles performances. It is detailed the facilities on post-irradiation examination. The techniques described are: online measurements, visual inspections, identifications of defective fuels and rods assemblies in spent fuel pools. This controls have made possible the feed-back to the manufactory process and the changes in the manufactory quality controls. (author)

A study on dissolution and leaching behaviour of spent nuclear fuels

International Nuclear Information System (INIS)

Lee, Chang Heon; Im, Hee Jung; Kim, Jong Gu; Park, Yang Soon; Ha, Yeong Keong

2010-12-01

This state of the art report describes a leaching behaviour of spent nuclear fuels which should be considered for safety assessment of spent nuclear fuel disposal in a deep geological repository. A decisive factor of a dissolution of UO 2 , a matrix of the fuel, is chemical characters (redox potential, pH, concentration of inorganic anions, water radiolysis subsequent by radiation field of the fuels) of ground water expected to be in contact with the fuels after the container has failed due to corrosion as well as atmosphere condition of a deep geological repository, which can change the oxidation state of UO 2 . The release rates of radionuclides from UO 2 matrix depend largely on their location within the fuels, that is, the radionuclides fixed in the fuel/cladding gap and grain boundaries are rapidly released. However, the radionuclides within the grains of the fuel are slowly released, and then their release rate is governed by a dissolution behaviour of UO 2

Automatic spent fuel ID number reader (I)

International Nuclear Information System (INIS)

Tanabe, S.; Kawamoto, H.; Fujimaki, K.; Kobe, A.

1991-01-01

An effective and efficient technique has been developed for facilitating identification works of LWR spent fuel stored in large scale spent fuel storage pools of such as processing plants. Experience shows that there are often difficulties in the implementation of operator's nuclear material accountancy and control works as well as safeguards inspections conducted on spent fuel assemblies stored in deep water pool. This paper reports that the technique is realized as an automatic spent fuel ID number reader system installed on fuel handling machine. The ID number reader system consists of an optical sub-system and an image processing sub-system. Thousands of spent fuel assemblies stored in under water open racks in each storage pool could be identified within relatively short time (e.g. within several hours) by using this combination. Various performance tests were carried out on image processing sub-system in 1990 using TV images obtained from different types of spent fuel assemblies stored in various storage pools of PWR and BWR power stations

Integrated quality status and inventory tracking system for FFTF driver fuel pins

International Nuclear Information System (INIS)

Gottschalk, G.P.

1979-11-01

An integrated system for quality status and inventory tracking of Fast Flux Test Facility (FFTF) driver fuel pins has been developed. Automated fuel pin identification systems, a distributed computer network, and a data base are used to implement the tracking system

Risk of transporting spent nuclear fuel by train

International Nuclear Information System (INIS)

Elder, H.K.

1981-12-01

This paper presents results of a study which analyzes the risk of transporting spent fuel by train. The risk assessment methodology consists of 4 basic steps: (1) a description of the system being analyzed; (2) identification of sequences of events that could lead to a release of material during transportation; (3) evaluation of the probability and consequences of each release sequence; and (4) assessment of the risk and evaluation of the results. The conclusion reached was that considering the substantial benefits derived from the fuel, the current spent fuel transportation system poses reasonably low risks

Characteristics of radioactive waste streams generated in HTGR fuel reprocessing

International Nuclear Information System (INIS)

Lin, K.H.

1976-01-01

Results are presented of a study concerned with identification and characterization of radioactive waste streams from an HTGR fuel reprocessing plant. Approximate quantities of individual waste streams as well as pertinent characteristics of selected streams have been estimated. Most of the waste streams are unique to HTGR fuel reprocessing. However, waste streams from the solvent extraction system and from the plant facilities do not differ greatly from the corresponding LWR fuel reprocessing wastes

An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

International Nuclear Information System (INIS)

Gillespie, G.E.

An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

The interpretation of fuel centre temperature measurements on a suspected leaking fuel pin

International Nuclear Information System (INIS)

Ainscough, J.B.; Lang, C.; Clough, D.J.

1983-01-01

In order to study fuel densification a series of single instrumented pin irradiations has been carried out in the High Pressure Water Loop of DIDO at Harwell. The behaviour of two of these pins was different from that expected. In the fifth test, where the fuel was 95% dense pellet UO 2 and expected to densify readily in-reactor, the fuel centre temperature increased from its starting value of approx. 1300 deg. C at a rate somewhat higher than expected on the basis of predicted densification rates. After about six days, the temperature increased rapidly and unexpectedly to 2100-2200 deg. C and remained steady at this level for a further eight days until a reactor trip occurred and the pin was unloaded. Predictions made using the HOTROD code imply a maximum fuel temperature of less than 1500 deg. C after densification. Post-irradiation examination confirmed that fission gas release had occurred, that the measured temperatures were consistent with the fuel microstructure and that the pin had a high internal gas pressure. The fourth pin in the series contained 97% dense UO 2 which was also expected to be dimensionally unstable. Qualitatively its behaviour was similar to that of the fifth pin though the temperatures throughout were lower. This pin experienced a number of major power cycles and failed after about 30 days in-reactor. It is probable that coolant ingress occurred in both pins via the thermocouple Hoke seal, degrading the filling gas conductivity and allowing the fuel to densify rapidly with consequent increase in the fuel/clad gap and hence in fuel temperature. These irradiations show that, for a short time at least, an apparently unfailed pin could operate undetected with temperatures significantly higher than those predicted for normal operation. (author)

Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors

International Nuclear Information System (INIS)

Sharaevsky, L.G.; Sharaevskaya, E.I.; Domashev, E.D.; Arkhypov, A.P.; Kolochko, V.N.

2002-01-01

The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on

Failing Decision

DEFF Research Database (Denmark)

Knudsen, Morten

2014-01-01

Recently the Danish subway trains have begun to announce “on time” when they arrive at a station on time. This action reflects a worrying acceptance of the normality of failure. If trains were generally expected to be on time, there would be no reason to – triumphantly – announce it. This chapter...... by an interest in failure as one way of improving understanding of present-day decision making in organizations.......Recently the Danish subway trains have begun to announce “on time” when they arrive at a station on time. This action reflects a worrying acceptance of the normality of failure. If trains were generally expected to be on time, there would be no reason to – triumphantly – announce it. This chapter...... deals not with traffic delays, but with failing decisions in organizations. The assumption of this chapter is that failing decisions today are as normal as delayed trains. Instead of being the exception, failure is part of the everyday reproduction of organizations – as an uncontrolled effect but also...

Simulating fail-stop in asynchronous distributed systems

Science.gov (United States)

Sabel, Laura; Marzullo, Keith

1994-01-01

The fail-stop failure model appears frequently in the distributed systems literature. However, in an asynchronous distributed system, the fail-stop model cannot be implemented. In particular, it is impossible to reliably detect crash failures in an asynchronous system. In this paper, we show that it is possible to specify and implement a failure model that is indistinguishable from the fail-stop model from the point of view of any process within an asynchronous system. We give necessary conditions for a failure model to be indistinguishable from the fail-stop model, and derive lower bounds on the amount of process replication needed to implement such a failure model. We present a simple one-round protocol for implementing one such failure model, which we call simulated fail-stop.

Nuclear Fuel Cycle System Analysis (I)

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Kwon, Eun Ha; Kim, Ho Dong; Yoon, Ji Sup; Park, Seong Won

2006-12-15

As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle, and evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance and economics. The analysis shows that the GEN-IV Recycle appears to have an advantage in terms of sustainability, environment-friendliness and long-term proliferation-resistance, while it is expected to be more economically competitive, if uranium ore prices increase or costs of pyroprocessing and fuel fabrication decrease.

Nitride fuels irradiation performance data base

International Nuclear Information System (INIS)

Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

1987-01-01

An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

Study on the behavior of unirradiated light water reactor fuel with iodine-127 under the reactivity initiated accident (RIA) conditions

International Nuclear Information System (INIS)

Sasajima, Hideo; Yanagisawa, Kazuaki; Kanazawa, Hiroyuki

1988-07-01

In a light water reactor fuel, a stress arised from pellet-cladding interaction (PCI) will have possibility to cause stress corrosion claddling (PCI failure) under an environment of corrosive fission product ; line iodine. A pulse irradiation experiment by NSRR was carried out to examine whether or not an unirradiated NSRR standard fuel rod in which 1.3 mg (33 x 10 -6 g/cm 2 ) of iodine was artificially filled could cause the PCI failure. Obtained results are: (1) The fuel rods with iodine did not fail both at deposited energy levels of 268 and 280 cal/g · UO 2 . On the other hand, the fuel rods without iodine failed at the same energy levels due to thinning of the cladding wall thickness. Within this experimental scope, PCI-failure did not occur on iodine filled fuel rods. (2) At a periphery of the fuel pellet of iodine filled rod, an uniform torus ring was formed. The torus ring consisted of an equi-axed large grains at 268 cal/g · UO 2 and a columnar ones at 280 cal/g · UO 2 . The torus ring was not formed in the fuel without iodine. (author)

Power ramping test in the JMTR for PCI study of water reactor fuel

International Nuclear Information System (INIS)

Nakata, H.; Kanbara, M.; Ichikawa, M.

1984-01-01

Power ramping test is essential for PCI study of water reactor fuel. Boiling water capsules have been used for the tests in the JMTR. Heat generation of fuel rod in the capsule can be changed by the He-3 power control facility during reactor operation. Four specially designed fuel rods have been ramped to about 41-43 kW/m; two of them have small gaps filled with iodine, the other two are equipped with centerline temperature thermocouple. Fuel rod elongation detector is equipped to each capsule. For the fuel rods with small gap, unique contraction followed by ordinary fuel relaxation behaviour was observed right after the fast ramping. None of them failed. Future programme includes a series of tests of fuel rods irradiated in the high-pressure water loop at the JMTR and a verification test of remedy fuel which allows daily-load-following operation of BWRs. (author)

Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

Energy Technology Data Exchange (ETDEWEB)

Carbajo, J.J.

2005-05-27

The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

Financial subsidies to the Australian fossil fuel industry

International Nuclear Information System (INIS)

Riedy, Chris; Diesendorf, Mark

2003-01-01

A common claim during international greenhouse gas reduction negotiations has been that domestic emissions cuts will harm national economies. This argument fails to consider the distorting effect of existing financial subsidies and associated incentives to fossil fuel production and consumption provided by governments in most developed countries. These subsidies support a fossil fuel energy sector that is the major contributor to global greenhouse gas emissions and conflict with attempts to expand the role of sustainable energy technologies. Reform of these types of subsidies has the potential to provide substantial gains in economic efficiency as well as reductions in carbon dioxide emissions--a 'no regrets' outcome for the economy and the environment. This paper examines financial subsidies to fossil fuel production and consumption in Australia and estimates the magnitude of the subsidies. Subsidies and associated incentives to fossil fuel production and consumption in Australia are similar to those in the United States and the other countries that have pushed for increased 'flexibility' during international negotiations

Spent fuel surveillance and monitoring methods

International Nuclear Information System (INIS)

1988-05-01

The Technical Committee Meeting on ''Spent Fuel Surveillance and Monitoring Methods'' (27-30 October 1987) has been organized in accordance with recommendations of the International Standing Advisory Group on Spent Fuel Management during its second meeting in 1986. The aim of the meeting was to discuss the above questions with emphasis on current design and operation criteria, safety principles and licensing requirements and procedures in order to prevent: inadvertent criticality, undue radiation exposure, unacceptable release of radioactivity as well as control for loss of storage pool water, crud impact, water chemistry, distribution and behaviour of particulates in cooling water, oxidation of intact and failed fuel rods as a function of temperature and burnup; distribution of radiation and temperature through dry cask wall, monitoring of leakages from pools and gas escapes from dry storage facilities, periodical integrity tests of the containment barriers, responsibilities of organizations for the required operation, structure, staff and subordination, etc. The presentations of the Meeting were divided into two sessions: Spent fuel surveillance programmes and practice in Member States (4 papers); Experimental methods developed in support of spent fuel surveillance programmes (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

Improvement of biogas as a fuel in molten carbonate fuel cells (MCFC); Aprovechamiento del biogas como combustible en pilas de combustible de carbonatos fundidos (MCFC)

Energy Technology Data Exchange (ETDEWEB)

Gil Diez, J.

2002-07-01

Molten carbonate fuel cells (MCFC) have a high efficiency of approx 50% when using biogas as a fuel and are among all types of FC the best suited for biogas. A precondition for use biogas in fuel cells is the reduction of accompanying traces of detrimental gases, therefore the RTD-work is two fold: A Preprocessing unit must be developed and the expected endurance must be confirmed. As a lesson learned in prior projects major reasons why renewable energy projects fail is the one-sided focus on technical aspects, that is why non-technical barriers shall be taken into account and realistic recommendations have to be established to overcome possible economic, logistic, legal and social problems. (Author)

Rule Based Reasoning Untuk Monitoring Distribusi Bahan Bakar Minyak Secara Online dan Realtime menggunakan Radio Frequency Identification

Directory of Open Access Journals (Sweden)

Mokhamad Iklil Mustofa

2017-05-01

Full Text Available The scarcity of fuel oil in Indonesia often occurs due to delays in delivery caused by natural factors or transportation constraints. Theaim of this  research is to develop systems of fuel distribution monitoring online and realtime using rule base reasoning method and radio frequency identification technology. The rule-based reasoning method is used as a rule-based reasoning model used for monitoring distribution and determine rule-based safety stock. The monitoring system program is run with a web-based computer application. Radio frequency identification technology is used by utilizing radio waves as an media identification. This technology is used as a system of tracking and gathering information from objects automatically. The research data uses data of delayed distribution of fuel from fuel terminal to consumer. The monitoring technique uses the time of departure, the estimated time to arrive, the route / route passed by a fuel tanker attached to the radio frequency Identification tag. This monitoring system is carried out by the radio frequency identification reader connected online at any gas station or specified position that has been designed with study case in Semarang. The results of the research covering  the status of rule based reasoning that sends status, that is timely and appropriate paths, timely and truncated pathways, late and on track, late and cut off, and tank lost. The monitoring system is also used in determining the safety stock warehouse, with the safety stock value determined based on the condition of the stock warehouse rules.

Introducing wood pellet fuel to the UK

Energy Technology Data Exchange (ETDEWEB)

Cotton, R A; Giffard, A

2001-07-01

Technical and non-technical issues affecting the introduction of wood pellet-fired heating to the UK were investigated with the aim of helping to establish a wood pellet industry in the UK. The project examined the growth and status of the industry in continental Europe and North America, reviewed relevant UK standards and legislation, identified markets for pellet heating in the UK, organised workshops and seminars to demonstrate pellet burning appliances, carried out a trial pelletisation of a range of biomass fuels, helped to set up demonstration installations of pellet-fired appliances, undertook a promotional campaign for wood pellet fuel and compiled resource directories for pellet fuel and pellet burning appliances in the UK. The work was completed in three phases - review, identification and commercialisation. Project outputs include UK voluntary standards for wood pellet fuel and combustion appliances, and a database of individuals with an interest in wood pellet fuel.

Method of detecting a fuel element failure

International Nuclear Information System (INIS)

Cohen, P.

1975-01-01

A method is described for detecting a fuel element failure in a liquid-sodium-cooled fast breeder reactor consisting of equilibrating a sample of the coolant with a molten salt consisting of a mixture of barium iodide and strontium iodide (or other iodides) whereby a large fraction of any radioactive iodine present in the liquid sodium coolant exchanges with the iodine present in the salt; separating the molten salt and sodium; if necessary, equilibrating the molten salt with nonradioactive sodium and separating the molten salt and sodium; and monitoring the molten salt for the presence of iodine, the presence of iodine indicating that the cladding of a fuel element has failed. (U.S.)

Utility residual fuel oil market conditions: An update

International Nuclear Information System (INIS)

Mueller, H.A. Jr.

1992-01-01

Planning for residual fuel oil usage and management remains an important part of the generation fuel planning and management function for many utilities. EPRI's Utility Planning Methods Center has maintained its analytical overview of the fuel oil markets as part of its overall fuel planning and management research program. This overview provides an update of recent fuel oil market directions. Several key events of the past year have had important implications for residual fuel oil markets. The key events have been the changes brought about by the Persian Gulf War and its aftermath, as well as continuing environmental policy developments. The Persian Gulf conflict has created renewed interest in reducing fuel oil use by utilities as part of an overall reduction in oil imports. The policy analysis performed to date has generally failed to properly evaluate utility industry capability. The Persian Gulf conflict has also resulted in an important change in the structure of international oil markets. The result of this policy-based change is likely to be a shift in oil pricing strategy. Finally, continued change in environmental requirements is continuing to shift utility residual oil requirements, but is also changing the nature of the US resid market itself

Status of the spent fuel dry storage programme for Cernavoda NPP

International Nuclear Information System (INIS)

Radu, M.

1999-01-01

The Cernavoda NPP Unit 1 (600 MWe Standard type) is in operation since December 1996. Within the framework of the R and D Radioactive Waste and Spent Fuel Management Programme, investigations, studies and research are carried out on site identification and conceptual designs for both a Spent Fuel Interim Storage Facility and a Spent Fuel Disposal Facility. The status of the work performed in the framework of this programme as well as the situation of the spent fuel resulting from the Research Institutes will be presented in the paper. (author)

Why did occidental modernity fail in the Arab Middle East: the failed modern state?

OpenAIRE

Sardar, Aziz

2011-01-01

This thesis asks a straightforward but nevertheless a complex question, that is: Why did modernity fail in the Arab Middle East? The notion of modernity in this thesis signifies the occidental modernity which reached the region in many different forms and through various channels. This occidental modernity had an impact on many areas and changed the societies and politics of the region. But these changes stopped short of reaching modernity, in other words it failed to change the society from ...

HTGR fuel behavior at very high temperature

International Nuclear Information System (INIS)

Kashimura, Satoru; Ogawa, Touru; Fukuda, Kousaku; Iwamoto, Kazumi

1986-03-01

Fuel behavior at very high temperature simulating abnormal transient of the reactor operation and accidents have been investigated on TRISO coating LEU oxide particle fuels at JAERI. The test simulating the abnormal transient was carried out by irradiation of loose coated particles above 1600 deg C. The irradiation test indicated that particle failure was principally caused by kernel migration. For simulation of the core heat-up accident, two experiments of out-of-pile heating were made. Survival temperature limits were measured and fuel performance at very high temperature were investigated by the heatings. Study on the fuel behavior under reactivity initiated accident was made by NSRR(Nuclear Safety Research Reactor) pulse irradiation, where maximum temperature was higher than 2800 deg C. It was found in the pulse irradiation experiments that the coated particles incorporated in the compacts did not so severely fail unlike the loose coated particles at ultra high temperature above 2800 deg C. In the former particles UO 2 material at the center of the kernel vaporized, leaving a spherical void. (author)

Application of eddy currents to post-irradiation examination of fuel rods

International Nuclear Information System (INIS)

Domizzi, G.; Ruch, M.; Ruggirello, G.; Spinosa, C.

1997-01-01

Postirradiation tests are performed on the fuel bundles of nuclear power plants, in order to evaluate their performance. The Zircaloy-4 cladding, the first containment of the fission products, is a very important part of these bundles. A fundamental step of these tests is the in-pool identification of the failed bars in the 'suspect' bundles. Later, once in the hot cell facility and prior to the destructive tests, it is necessary to characterize the defects in the cladding. The eddy current method provides a means for fast and reliable detection and characterisation of defects unobservable in visual inspection, such as tiny cracks, pores and anomalously hydride regions. The project for the application of this method in postirradiation tests has been divided into three stages, namely laboratory set up, in-pool tests, hot-cell application, the first one being described here. Techniques for the construction of synthetic defects (machined, micro cracks, abnormal hydride concentration, hydride blisters, oxide layers) were developed. A mechanical device for automatic probe movement was designed and constructed. Special external probes for the particular defects were developed. The inspection procedure was prepared. (author) [es

Analysis of possibilities for functional capacity for work rise of reactor fuel elements at nuclear engine regime

International Nuclear Information System (INIS)

Deryavko, I.I.; Perepelkin, I.G.; Pivovarov, O.S.; Storozhenko, A.N.; Tarasov, V.I.

2000-01-01

The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

Accurate Identification of Cancerlectins through Hybrid Machine Learning Technology.

Science.gov (United States)

Zhang, Jieru; Ju, Ying; Lu, Huijuan; Xuan, Ping; Zou, Quan

2016-01-01

Cancerlectins are cancer-related proteins that function as lectins. They have been identified through computational identification techniques, but these techniques have sometimes failed to identify proteins because of sequence diversity among the cancerlectins. Advanced machine learning identification methods, such as support vector machine and basic sequence features (n-gram), have also been used to identify cancerlectins. In this study, various protein fingerprint features and advanced classifiers, including ensemble learning techniques, were utilized to identify this group of proteins. We improved the prediction accuracy of the original feature extraction methods and classification algorithms by more than 10% on average. Our work provides a basis for the computational identification of cancerlectins and reveals the power of hybrid machine learning techniques in computational proteomics.

Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1992-03-01

The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

WWER expert system for fuel failure analysis using the RTOP-CA code

International Nuclear Information System (INIS)

Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

2008-01-01

The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

California's Low-Carbon Fuel Standard - Compliance Trends

Science.gov (United States)

Witcover, J.; Yeh, S.

2013-12-01

Policies to incentivize lower carbon transport fuels have become more prevalent even as they spark heated debate over their cost and feasibility. California's approach - performance-based regulation called the Low Carbon Fuel Standard (LCFS) - has proved no exception. The LCFS aims to achieve 10% reductions in state transport fuel carbon intensity (CI) by 2020, by setting declining annual CI targets, and rewarding fuels for incremental improvements in CI beyond the targets while penalizing those that fail to meet requirements. Even as debate continues over when new, lower carbon fuels will become widely available at commercial scale, California's transport energy mix is shifting in gradual but noticeable ways under the LCFS. We analyze the changes using available data on LCFS fuels from the California Air Resources Board and other secondary sources, beginning in 2011 (the first compliance year). We examine trends in program compliance (evaluated through carbon credits and deficits generated), and relative importance of various transport energy pathways (fuel types and feedstocks, and their CI ratings, including new pathways added since the program's start). We document a roughly 2% decline in CI for gasoline and diesel substitutes under the program, with compliance achieved through small shifts toward greater reliance on fuels with lower CI ratings within a relatively stable amount of transport energy derived from alternatives to fossil fuel gasoline and diesel. We also discuss price trends in the nascent LCFS credit market. The results are important to the broader policy debate about transportation sector response to market-based policies aimed at reducing the sector's greenhouse gas emissions.

Solid oxide fuel cell field trial evaluation

Energy Technology Data Exchange (ETDEWEB)

Wilcox, C.P.; Winstanley, R.; Nietsch, T.; Smith, C.; Knight, R.; Seymore, C.

2000-07-01

This report focuses on issues relating to a field trial of a solid oxide fuel cell (SOFC). Aspects examined include markets for SOFC systems, the choice of systems for demonstration in year 2002, the assessment of industrial interest, and evaluation and ranking of candidate systems. The identification and evaluation of interest in field trials, the estimation of the capital and running costs of a field trial, and identification of the benefits to the UK and barriers to implementation of SOFC systems are discussed. (UK)

Closed External Fixation for Failing or Failed Femoral Shaft Plating in a Developing Country.

Science.gov (United States)

Aliakbar, Adil; Witwit, Ibrahim; Al-Algawy, Alaa A Hussein

2017-08-01

Femoral shaft fractures are one of the common injuries that is treated by open reduction, with internal fixation by plate and screws or intramedullary nailing, which can achieve a high union rate. To evaluate the outcome of using closed external fixation to augment a failing plate; with signs of screw loosening and increasing bone/plate gap; a failed plate; broken plate; screws completely out of bone with redisplacement of fracture. A retrospective study on 18 patients, aged between 17-42 years, who presented between 6-18 weeks after initial surgical fixation, with pain, difficulty in limb function, deformity and abnormal movement at fracture site, was done. X-Rays showed plating failure with acceptable amount of callus, which unfortunately had refractured. Cases associated with infection and no radiological evidence of callus formation were excluded from this study. Closed reduction was done by manipulation, then fracture fixation by AO external fixator. The patients were encouraged for full weight bearing as early as possible with dynamization later on. Of the 18 patients who underwent external fixation after close reduction, 15 cases showed bone healing in a period between 11-18 weeks (mean of 14.27 weeks) with good alignment (Radiologically). Removal of external fixator was done followed by physical therapy thereafter. Closed external fixation for treatment of failing or failed femoral plating, achieves good success rate and has less complications, is a short time procedure, especially in a hospital with limited resources.

Plan of studies on fuel failure detection in Rikkyo Research Reactor

International Nuclear Information System (INIS)

Matsuura, T.; Nagahara, T.; Hattori, M.; Kawaguchi, K.

1987-01-01

Studies on fuel failure detection in Rikkyo Research Reactor have recently been begun in the following four approaches. (1) Accumulation of the data on the concentration of the short-lived radioactivity originating from FP rare gases contained in the air on the water surface of the reactor tank. (2) Accumulation of the data on the concentration of FP (especially 131 I) in the water of the reactor tank. (3) Design and preparation of a ''sniffer'' by which the location of the failed fuel element can be detected, when some anomaly is found in the above two routine measurements. (4) Design and preparation of a vessel containing a fuel element, which can be useful both for ''sipping'' inspection of the fuel element and for storage of the damaged fuel element. In this paper, an outline of the above approaches and the results of some preliminary experiments are reported. (author)

Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

International Nuclear Information System (INIS)

Thoms, K.R.

1975-04-01

Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 (Ta-8 percent W-2 percent Hf) and uranium dioxide (UO 2 ) fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1 percent Zr. A total of nine fuel pins was irradiated (four containing porous UN, two containing dense, nonporous UN, and three containing dense UO 2 ) at average cladding temperatures ranging from 931 to 1015 0 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO 2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of 10 -5 . (U.S.)

Progress of fossil fuel science

Energy Technology Data Exchange (ETDEWEB)

Demirbas, M.F.

2007-07-01

Coal is the most abundant and widely distributed fossil fuel. More than 45% of the world's electricity is generated from coal, and it is the major fuel for generating electricity worldwide. The known coal reserves in the world are enough for more than 215 years of consumption, while the known oil reserves are only about 39 times of the world's consumption and the known natural gas reserves are about 63 times of the world's consumption level in 1998. In recent years, there have been effective scientific investigations on Turkish fossil fuels, which are considerable focused on coal resources. Coal is a major fossil fuel source for Turkey. Turkish coal consumption has been stable over the past decade and currently accounts for about 24% of the country's total energy consumption. Lignite coal has had the biggest share in total fossil fuel production, at 43%, in Turkey. Turkish researchers may investigate ten broad pathways of coal species upgrading, such as desulfurization and oxydesulfurization, pyrolysis and hydropyrolysis, liquefaction and hydroliquefaction, extraction and supercritical fluid extraction, gasification, oxidation, briquetting, flotation, and structure identification.

Information technology in fuel manufacturing

International Nuclear Information System (INIS)

Seshagiri Rao, G.R.; Arora, U.K.; Mohanty, Deepak; Siva Kumar, G.V.S.M.; Banerjee, P.K.

2012-01-01

NFC, Hyderabad is engaged in manufacturing of fuel assemblies required for Indian Nuclear Power Programme. During the manufacturing process, the basic Uranium Fuel and Zirconium alloy cladding tubes travels through several work centers, machines and exposes to various process parameters. For analyzing the fuel performance these parameters are indicators and is a requirement to record such history by both manufacturer and customer. NFC has planned to deploy Information Technology (IT) Systems from MDU/UOC Dissolution to Finished Fuel Assembly dispatch stage by using Radio Frequency IDentification (RFID)/Barcode Technologies. IT Systems are connected to electronic weigh balances to acquire material weight data automatically. The IT Systems are also designed to receive data from small Island Systems like Helium leak testing equipments. As a pilot project the system is initially implemented from empty tube Tray preparation stage to Fuel Assembly Packing and dispatch stage, containing about more than 14 processes. The system was built using open source technology platform and was deployed on a cost effective Hardware environment. The present paper describes the development process of the system, Implementation challenges faced and change management. The paper also discusses about fruits of implementation and productivity improvements. (author)

Evaluation of LWR fuel rod behavior under operational transient conditions

International Nuclear Information System (INIS)

Nakamura, M.; Hiramoto, K.; Maru, A.

1984-01-01

To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

Neglected City Narratives And Failed Rebranding

DEFF Research Database (Denmark)

Mousten, Birthe; Locmele, Gunta

2017-01-01

Rīga, Latvia went through a failed rebranding process as the forerunner of its status as a European Capital of Culture (2014). The same thing happened in Aarhus, Denmark. Aarhus will be a European Capital of Culture (2017) and leading to this, it went through a failed rebranding process. Based on...

When Organization Fails: Why Authority Matters

DEFF Research Database (Denmark)

Blaschke, Steffen

2015-01-01

Review of: James R. Taylor and Elizabeth J. Van Every / When Organization Fails: Why Authority Matters. (New York: Routledge, 2014. 220 pp. ISBN: 978 0415741668)......Review of: James R. Taylor and Elizabeth J. Van Every / When Organization Fails: Why Authority Matters. (New York: Routledge, 2014. 220 pp. ISBN: 978 0415741668)...

The cost of fuel economy in the Indian passenger vehicle market

International Nuclear Information System (INIS)

Chugh, Randy; Cropper, Maureen; Narain, Urvashi

2011-01-01

To investigate how fuel economy is valued in the Indian car market, we compute the cost to Indian consumers of purchasing a more fuel-efficient vehicle and compare it to the benefit of lower fuel costs over the life of the vehicle. We estimate hedonic price functions for four market segments (petrol hatchbacks, diesel hatchbacks, petrol sedans, and diesel sedans) to compute 95% confidence intervals for the marginal cost to the consumer for an increase in fuel economy. We find that the associated present value of fuel savings falls within the 95% confidence interval for most specifications, in all market segments, for the years 2002 through 2006. Thus, we fail to consistently reject the hypothesis that consumers appropriately value fuel economy. - Highlights: → We examine the tradeoffs faced by new vehicle consumers in India. → We use hedonic price functions and instrumental variables. → We find no support for the hypothesis that consumers undervalue fuel economy. → Some consumers are willing to forgo substantial potential savings to own their preferred vehicle.

Regulatory experience with fuel failures in Switzerland

International Nuclear Information System (INIS)

Adam, L.

2015-01-01

In this paper the main ENSI activities like: supervision of reactor and radiation safety and security; supervision of safety of transports of nuclear materials and assess the safety of proposed solutions for the geological disposal are listed. Recent events concerning the reactor core, common causes for fuel failures, findings during inspections and potential root cause for fuel failures are discussed. Management of fuel failures, started from reporting of the event – evaluation of the need of imminent action; identification of the fuel element if possible till evaluation by the plant and fuel vendor and allowance by ENSI for repair of the fuel element and definition of measures (short and long term) are also presented. The following Conclusions by ENSI about status of fuel failures are made: 1) Number of fuel failures was reduced regardless more economic operation in all plants; 2) Old PWR and BWR reactors achieved 15 to 29 years operation without leakers, but two minor fuel damage during fuel handling appeared; 3) Newer plants are not better in achieving operation without leakers than older plants; 4) Technical improvements at fuel elements parallel to changes in operation strategy and improvements in manufacturing quality but single effects difficult to judge. The issues about how to implement “Zero Failure Rates” in regulations and how to achieve “Zero Failure Rates” as well as some future measures by ENSI are discussed

MRI as a problem-solving tool in unexplained failed total hip replacement following conventional assessment

International Nuclear Information System (INIS)

Johnston, Ciaran; Kerr, Jennifer; Ford, Stephanie; Eustace, Stephen; O'Byrne, John

2007-01-01

To evaluate MRI as a problem-solving tool for patients with an unexplained failed total hip replacement following conventional radiological assessment. Patients' informed consent was obtained in all cases. Institutional review board approval was obtained. Twenty-eight patients with unexplained failed total hip replacements following conventional radiological assessment underwent additional MR imaging with an optimised turbo-spin echo sequence. Images were reviewed by two musculoskeletal radiologists by consensus and compared with findings at surgery, or following response to image-guided intervention or clinical follow-up. Of the 28 patients, MRI revealed an unsuspected diagnosis explaining the cause of prosthesis failure in 15 patients. In eight of 15 patients in this group, subsequent minimally invasive image-guided intervention obviated the need for revision total hip replacement. No cause for prosthesis failure was identified in 13 patients. MRI may be successfully undertaken in patients following total hip replacement, and, when performed, it frequently leads to an unsuspected diagnosis, allowing informed patient treatment. In this study it allowed the identification of an unsuspected diagnosis in over 50% of cases. (orig.)

Strength of Anode‐Supported Solid Oxide Fuel Cells

DEFF Research Database (Denmark)

Faes, A.; Frandsen, Henrik Lund; Kaiser, Andreas

2011-01-01

Nickel oxide and yttria doped zirconia composite strength is crucial for anode‐supported solid oxide fuel cells, especially during transient operation, but also for the initial stacking process, where cell curvature after sintering can cause problems. This work first compares tensile and ball....... Even though the electrolyte is to the tensile side, it is found that the anode support fails due to the thermo‐mechanical residual stresses....

Incorporation of new tables into the preliminary structure of BDCN - nuclear fuel data base

International Nuclear Information System (INIS)

Pombo, J.B.S.; Andrade, M.C.

1989-01-01

The structure of the new tables related to fuel elements, fuel rods and fuel pellets have been incorporated into the developing structure of BDCN - nuclear fuel data base, are described. The list of all 44 tables structured up to now, with identification of its primary keys, as well as the reduced set of data has been entered to run the simulations are also presented. The simulation tests have been ran using a microcomputer.(M.I.)

7 CFR 983.52 - Failed lots/rework procedure.

Science.gov (United States)

2010-01-01

... 7 Agriculture 8 2010-01-01 2010-01-01 false Failed lots/rework procedure. 983.52 Section 983.52..., ARIZONA, AND NEW MEXICO Regulations § 983.52 Failed lots/rework procedure. (a) Substandard pistachios... committee may establish, with the Secretary's approval, appropriate rework procedures. (b) Failed lot...

Dynamics of Practical Premixed Flames, Part I: Model Structure and Identification

Directory of Open Access Journals (Sweden)

A. Huber

2009-06-01

Full Text Available For the analysis of thermoacoustic instabilities it is most important to determine the dynamic flame response to acoustic disturbances. Premixed flames are often modelled as single-input single-output system, where the “output” (the overall rate of heat release responds to a single “input” variable (often the velocity at the exit of the burner nozzle. However, for practical premixed flames, where perturbations of pressure or velocity at the fuel injector will modulate the fuel equivalence ratio, the heat release rate will respond to fluctuations of equivalence ratio as well as nozzle mass flow rate. In this case, a multiple-input, single-output (MISO model structure for the flame is appropriate. Such a model structure is developed in the present paper. Staged fuel injection as well as fuel line impedances can be taken into account, the integration with low-order or finite-element based models for stability analysis is straightforward. In order to determine unit impulse and frequency response functions for such a model structure, an identification scheme based on unsteady CFD calculation with broadband excitation followed by correlation analysis is proposed and validated successfully. Identification of MISO model coefficients is a challenging task, especially in the presence of noise. Therefore criteria are introduced which allow to ascertain a posteriori how well the identified model represents the true system dynamics. Using these criteria, it is investigated how excitation signal type, time series length and signal-to-noise ratio influence the results of the identification process. Consequences for passive design strategies based on multi-stage fuel injection and experimental work on practical premixed flame dynamics are discussed.

Acceptance of non-fuel assembly hardware by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high-level waste will be accepted in the following categories: failed fuel; consolidated fuel and associated structural parts; non-fuel-assembly hardware; fuel in metal storage casks; fuel in multi-element sealed canisters; inspection and testing requirements for wastes; canister criteria; spent fuel selection for delivery; and defense and commercial high-level waste packages. 14 refs., 12 figs., 43 tabs

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

Energy Technology Data Exchange (ETDEWEB)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

2016-10-15

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

International Nuclear Information System (INIS)

Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae

2016-01-01

Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed

Quality analysis in pressurized water reactor fuel

International Nuclear Information System (INIS)

Darolles, J.F.

1975-01-01

An integrated system which has been set up to administrate and analyze the quality is described. This system is in actual operation. The basic principles for quality analysis system are traceability, i.e., identification, location and history of fuel components and quality evaluation during manufacturing. The quality analysis system operates in the following areas: data recording and transmission, data processing, quality file generation. The interest of such a system may be noted particularly in manufacturing, for the constitution of quality files, the design of products and the processing of data from irradiated fuel assemblies [fr

DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Miles, K.J.; Hill, D.J.

1986-01-01

The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

Mixed fuel strategy for carbon deposition mitigation in solid oxide fuel cells at intermediate temperatures.

Science.gov (United States)

Su, Chao; Chen, Yubo; Wang, Wei; Ran, Ran; Shao, Zongping; Diniz da Costa, João C; Liu, Shaomin

2014-06-17

In this study, we propose and experimentally verified that methane and formic acid mixed fuel can be employed to sustain solid oxide fuel cells (SOFCs) to deliver high power outputs at intermediate temperatures and simultaneously reduce the coke formation over the anode catalyst. In this SOFC system, methane itself was one part of the fuel, but it also played as the carrier gas to deliver the formic acid to reach the anode chamber. On the other hand, the products from the thermal decomposition of formic acid helped to reduce the carbon deposition from methane cracking. In order to clarify the reaction pathways for carbon formation and elimination occurring in the anode chamber during the SOFC operation, O2-TPO and SEM analysis were carried out together with the theoretical calculation. Electrochemical tests demonstrated that stable and high power output at an intermediate temperature range was well-maintained with a peak power density of 1061 mW cm(-2) at 750 °C. With the synergic functions provided by the mixed fuel, the SOFC was running for 3 days without any sign of cell performance decay. In sharp contrast, fuelled by pure methane and tested at similar conditions, the SOFC immediately failed after running for only 30 min due to significant carbon deposition. This work opens a new way for SOFC to conquer the annoying problem of carbon deposition just by properly selecting the fuel components to realize their synergic effects.

Fuel failure in water reactors: Causes and mitigation. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2003-03-01

The objective of this technical meeting (TM) was to review the present knowledge of the causes and mechanisms of fuel failure in water reactors during normal operational conditions. Emphasis has been given to analysis of failure causes and their mitigation by means of design as well as plant and core operation including strategies for operation with failed fuel. Some information on detection techniques (on-line monitoring and diagnostics, flux tilting, sipping techniques, etc) has also been presented. This TM presented also the progress on the above-mentioned subjects since the last meeting held in 1992 (Dimitrovgrad, Russian Federation). The topics covered in the papers were as follows: Experience feedback on fuel reliability (8 papers); Strategies to avoid or mitigate fuel failures (4 papers); Experimental studies on fuel failures and degradation mechanisms (4 papers); Modelling of fuel failure mechanisms (3 papers); Detection and monitoring during operation or outage (4 papers); Modelling and assessment of fuel failures (3 papers)

Irradiation performance of AGR-1 high temperature reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

2016-09-15

Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

International Nuclear Information System (INIS)

Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

1977-01-01

The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

Nuclear fuel manufacturing. Current activities and prospects at INR Pitesti

International Nuclear Information System (INIS)

Horhoianu, Grigore

2001-01-01

Development of the CANDU nuclear fuel is currently conducted world wide onto two principal directions: - increasing the service span of the current type of fuel and improving the efficiency of burnup in reactor; - reducing the costs of fuel manufacturing by improving the design and manufacturing technologies in condition of increasing fuel performance. In parallel, a research program, RAAN, is undergoing, concerning the development of advanced CANDU type fuels (SEU, RU, DUPIC, Th), aiming at reducing the overall costs per fuel cycle. In the INR TRIGA reactor a large number of experimental fuel elements manufactured in INR were irradiated under different conditions specific to the CANDU reactor operation. Post irradiation investigations both destructive and non-destructive were carried out in the hot cells at INR Pitesti. The experimental results were used in order to optimize and evaluate the fuel project, to check the fuel manufacturing technologies as well as to certify the computational codes. The local thermo-mechanical analyses by final element methods, modelling the SCC phenomenon, probabilistic evaluation of performance parameters of the fuel, constitute new directions in the modelling and developing computational code. The developed codes were submitted to a thorough validation process to comply with the quality assurance. The excellent results obtained in INR were confirmed by participation in the FUMEX International Exercises of computer code intercomparison, organized by IAEA Vienna. Progress was also recorded in establishing the behaviour of fuel elements failed during reactor operation and the effect their maintenance in the reactor core could have upon the power reactor operation. A system-expert variant was worked out able for a short term analysis of the decisions referring to removing the failing element at Cernavoda NPP. As advanced CANDU fuel is concerned, until now preliminary variants for a fuel bundle with 43 elements containing slightly

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

Energy Technology Data Exchange (ETDEWEB)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

International Nuclear Information System (INIS)

Schneider, K.J.; Kabele, T.J.

1979-09-01

To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems

Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

International Nuclear Information System (INIS)

Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

1990-05-01

The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

Safeguards and nonproliferation aspects of a dry fuel recycling technology

International Nuclear Information System (INIS)

Pillay, K.K.S.

1993-01-01

Los Alamos National Laboratory undertook an independent assessment of the proliferation potentials and safeguardability of a dry fuel recycling technology, whereby spent pressurized-water reactor (PWR) fuels are used to fuel canadian deuterium uranium (CANDU) reactors. Objectives of this study included (1) the evaluation of presently available technologies that may be useful to safeguard technology options for dry fuel recycling (2) and identification of near-term and long-term research needs to develop process-specific safeguards requirements. The primary conclusion of this assessment is that like all other fuel cycle alternatives proposed in the past, the dry fuel recycle entails prolfferation risks and that there are no absolute technical fixes to eliminate such risks. This study further concludes that the proliferation risks of dry fuel recycling options are relatively minimal and presently known safeguards systems and technologies can be modified and/or adapted to meet the requirements of safeguarding such fuel recycle facilities

Fuel cells for portable, mobile and hybrid applications

International Nuclear Information System (INIS)

Roberge, R.; Kaufman, A.

2002-01-01

The introduction of fuel cell systems for a variety of low-power applications (below 1000 watts) means they can be used for applications such as portable power sources and mobile power sources. The energy and power are separate elements in a fuel cell system. The power is provided by the fuel cell stack (output characteristics are dependent on the cell active area, number of cells, and operating conditions), and the energy is defined by the fuel (hydrogen) storage. The authors indicated that proton exchange membrane fuel cells are the most appropriate for small fuel cell systems, since they have a temperature range ambient to 90 Celsius, ambient air (non-humidified), and load following response. In addition, they possess a solid electrolyte, high power density and specific power, and low-pressure operation. Simplicity of operation is the key to the design of a fuel cell system. The parameters to be considered include hydrogen supply, air supply, water management, and thermal management. Some of the options available for fuels are: compressed hydrogen, metal hydrides, chemical hydrides, and carbon-based hydrogen storage. Some of the factors that will help in determining market penetration are: rapid cost reduction with volume, fuel infrastructure, proven reliability, and identification of applications where fuel cells provide superior performance. 2 figs

Households facing constraints. Fuel poverty put into context

International Nuclear Information System (INIS)

Dubois, Ute; Meier, Helena

2014-01-01

The present paper discusses the concept of fuel poverty taking into account the arbitrages made by households when they are facing economic constraints. Fuel poverty is still lacking a common definition throughout Europe: while the UK and France have (different) official definitions, there is still no definition in a country like Germany, or at the European level. Where definitions exist, they often consider that fuel poor households have high energy needs. The possibility of being fuel poor even without having high energy needs and the various arbitrage possibilities of households - i.e. to under-spend and use too little energy - are not systematically discussed. Our paper tries to fill that gap by putting fuel poverty into the larger context of constraints faced by households. Based on a graphical analysis, it shows that different situations of fuel poverty might occur. It results in the identification of two distinct fuel poverty problems: an ''energy inequality'' problem, reflected by the fact that some households pay disproportionately high energy bills, and an ''energy affordability'' problem that can affect a larger share of the population. It finally explores the two types of fuel poverty for European countries and discusses policy implications.

Characteristics of fuel cycle waste

International Nuclear Information System (INIS)

Aquilina, C.A.; Everette, S.E.

1982-01-01

The Low-Level Waste Management System started in 1979 to describe and model the existing commercial low-level waste management system. The system description produced is based on the identification of the different elements making up both the fuel and non-fuel cycle and their relationships to each other. A systems model based on the system description can accurately reflect the flow of low-level waste from generator to disposal site and is only limited by the reliability of the information it uses. For both the fuel cycle and non-fuel cycle large quantities of information is required in order to allow the system to operate at its full potential. Work is ongoing to collect this information. Significant progress is being made in the fuel cycle area primarily because the majority of fuel cycle low-level radioactive waste is produced by commercial power reactor plant operations. The Low-Level Waste Management system is only beginning to derive the benefits to be obtained from an accurate low-level waste management information system. As data is verified and analyzed, results on a national as well as individual organization level will be gained. Comparisons to previous studies will be made. Accurate projections of waste volumes and activities to be produced, projected impacts of waste streams of treatment or management changes are only examples of information to be produced. 1 figure, 1 table

Spent reactor fuel benchmark composition data for code validation

International Nuclear Information System (INIS)

Bierman, S.R.

1991-09-01

To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays are being obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Destructive assay data are being obtained from representative reactor fuels having experienced irradiation exposures up to about 55 GWD/MTM. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of the same fuel rod and represent radiation exposures of about 27, 37, and 44 GWD/MTM. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input

Palliative effects of H2 on SOFCs operating with carbon containing fuels

Science.gov (United States)

Reeping, Kyle W.; Bohn, Jessie M.; Walker, Robert A.

2017-12-01

Chlorine can accelerate degradation of solid oxide fuel cell (SOFC) Ni-based anodes operating on carbon containing fuels through several different mechanisms. However, supplementing the fuel with a small percentage of excess molecular hydrogen effectively masks the degradation to the catalytic activity of the Ni and carbon fuel cracking reaction reactions. Experiments described in this work explore the chemistry behind the "palliative" effect of hydrogen on SOFCs operating with chlorine-contaminated, carbon-containing fuels using a suite of independent, complementary techniques. Operando Raman spectroscopy is used to monitor carbon accumulation and, by inference, Ni catalytic activity while electrochemical techniques including electrochemical impedance spectroscopy and voltammetry are used to monitor overall cell performance. Briefly, hydrogen not only completely hides degradation observed with chlorine-contaminated carbon-containing fuels, but also actively removes adsorbed chlorine from the surface of the Ni, allowing for the methane cracking reaction to continue, albeit at a slower rate. When hydrogen is removed from the fuel stream the cell fails immediately due to chlorine occupation of methane/biogas reaction sites.

Development of treatment technology of radio-contaminated coolant in fuel test loop

International Nuclear Information System (INIS)

Kim, J. Y.

1997-10-01

In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs

Development of treatment technology of radio-contaminated coolant in fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.

1997-10-01

In 1995, the installation of KMRR located in KAERI provided a milestone in independence of nuclear technologies in Korea. The independence of technologies is only possible through the enormous investment for research and through the active approaches for various experiments. The performance of various experiments enhanced the risk of environmental pollution and the nuclear fuel irradiation test is one of those experiments. The damage of fuel which might happen any time in irradiation test, will discharge high level radioactive materials from the inside of failed fuel and will gradually contaminate the cooling water in near vicinity. Accordingly, the proper management of coolant having high temperature and high level . (author). refs., tabs., figs.

Proton Exchange Membrane Fuel Cell Modelling Using Moving Least Squares Technique

Directory of Open Access Journals (Sweden)

Radu Tirnovan

2009-07-01

Full Text Available Proton exchange membrane fuel cell, with low polluting emissions, is a great alternative to replace the traditional electrical power sources for automotive applications or for small stationary consumers. This paper presents a numerical method, for the fuel cell modelling, based on moving least squares (MLS. Experimental data have been used for developing an approximated model of the PEMFC function of the current density, air inlet pressure and operating temperature of the fuel cell. The method can be applied for modelling others fuel cell sub-systems, such as the compressor. The method can be used for off-line or on-line identification of the PEMFC stack.

Selection and use of a low enriched fuel in high performance research reactors

International Nuclear Information System (INIS)

Cerles, J.M.; Schwartz, J.P.

1978-08-01

A new nuclear fuel composition for research reactors (Osiris, Siloe) is studied using low enriched (E<20%) uranium oxide. Its utilization leads to modifications in the facilities of these experimental reactors: increase of primary coolant flow, modifications in failed element detection system, handling of materials and storage

New Optical Sensor Suite for Ultrahigh Temperature Fossil Fuel Application

Energy Technology Data Exchange (ETDEWEB)

John Coggin; Tom Flynn; Jonas Ivasauskas; Daniel Kominsky; Carrie Kozikowski; Russell May; Michael Miller; Tony Peng; Gary Pickrell; Raymond Rumpf; Kelly Stinson-Bagby; Dan Thorsen; Rena Wilson

2007-12-31

Accomplishments of a program to develop and demonstrate photonic sensor technology for the instrumentation of advanced powerplants and solid oxide fuel cells are described. The goal of this project is the research and development of advanced, robust photonic sensors based on improved sapphire optical waveguides, and the identification and demonstration of applications of the new sensors in advanced fossil fuel power plants, where the new technology will contribute to improvements in process control and monitoring.

Uranium dioxide Caramel fuel

International Nuclear Information System (INIS)

Schwartz, J.P.

The work performed in France on Caramel fuels for research reactors reflects the reality of a program based on non proliferation criteria, as they have already appeared several years ago. This work actually includes the following different aspects: identification of the non proliferation criterion defining this action; determination of the economical and technical goals to be reached; realization of research and development studies finalized in a full scale demonstration; transposition to an industrial and commercial level

Parameter Estimation of Spacecraft Fuel Slosh Model

Science.gov (United States)

Gangadharan, Sathya; Sudermann, James; Marlowe, Andrea; Njengam Charles

2004-01-01

Fuel slosh in the upper stages of a spinning spacecraft during launch has been a long standing concern for the success of a space mission. Energy loss through the movement of the liquid fuel in the fuel tank affects the gyroscopic stability of the spacecraft and leads to nutation (wobble) which can cause devastating control issues. The rate at which nutation develops (defined by Nutation Time Constant (NTC can be tedious to calculate and largely inaccurate if done during the early stages of spacecraft design. Pure analytical means of predicting the influence of onboard liquids have generally failed. A strong need exists to identify and model the conditions of resonance between nutation motion and liquid modes and to understand the general characteristics of the liquid motion that causes the problem in spinning spacecraft. A 3-D computerized model of the fuel slosh that accounts for any resonant modes found in the experimental testing will allow for increased accuracy in the overall modeling process. Development of a more accurate model of the fuel slosh currently lies in a more generalized 3-D computerized model incorporating masses, springs and dampers. Parameters describing the model include the inertia tensor of the fuel, spring constants, and damper coefficients. Refinement and understanding the effects of these parameters allow for a more accurate simulation of fuel slosh. The current research will focus on developing models of different complexity and estimating the model parameters that will ultimately provide a more realistic prediction of Nutation Time Constant obtained through simulation.

In-pile study of the reaction between breeder fuel and sodium

International Nuclear Information System (INIS)

Hugot, J.P.

1982-10-01

Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

WWER problems and perspectives of moving toward zero fuel failures and mitigation of fuel failure consequences at NPPs with WWER reactors in Russia

International Nuclear Information System (INIS)

Shestakov, Yu.; Semenovykh, A.

2015-01-01

The paper contained information on mitigation of fuel failures and perspectives of moving to ‘Zero Failure Level’. It was stated that driving to ‘zero’ failure rate should include two kinds of efforts, 1) focus on identification of failure mechanisms, and 2) implement corrective actions. It is also important to mitigate the consequences of fuel failures if that inevitably occurs. Fuel failures entail the risks of severe secondary degradation and contamination of primary circuit due to fuel washout. Significant changes of fuel operating conditions (longer fuel cycles, higher fuel burnup, power uprate) and innovations in fuel design bear the risk of higher failure rates for some period of time. Simultaneous implementation of several innovations in one nuclear utility is not advisable since it might be difficult to identify which of the innovations affected the fuel performance positively or negatively. The disadvantage of gradual implementation of corrective actions and any significant changes in operating conditions is the long time needed to evaluate the operational experience. In some cases, it may take up to 4-6 years for each significant change to reflect on the operational performance

Internal combustion engines for alcohol motor fuels: a compilation of background technical information

Energy Technology Data Exchange (ETDEWEB)

Blaser, Richard

1980-11-01

This compilation, a draft training manual containing technical background information on internal combustion engines and alcohol motor fuel technologies, is presented in 3 parts. The first is a compilation of facts from the state of the art on internal combustion engine fuels and their characteristics and requisites and provides an overview of fuel sources, fuels technology and future projections for availability and alternatives. Part two compiles facts about alcohol chemistry, alcohol identification, production, and use, examines ethanol as spirit and as fuel, and provides an overview of modern evaluation of alcohols as motor fuels and of the characteristics of alcohol fuels. The final section compiles cross references on the handling and combustion of fuels for I.C. engines, presents basic evaluations of events leading to the use of alcohols as motor fuels, reviews current applications of alcohols as motor fuels, describes the formulation of alcohol fuels for engines and engine and fuel handling hardware modifications for using alcohol fuels, and introduces the multifuel engines concept. (LCL)

Pass/fail patterns of candidates who failed COMLEX-USA level 2-PE because of misrepresentation of clinical findings on postencounter notes.

Science.gov (United States)

Langenau, Erik E; Sandella, Jeanne M

2011-07-01

In 2007, The National Board of Osteopathic Medical Examiners (NBOME) instituted a policy to address the accuracy and integrity of postencounter written documentation recorded during the Comprehensive Osteopathic Medical Licensing Examination Level 2-Performance Evaluation (COMLEX-USA Level 2-PE). This policy was instituted not only to protect the integrity of the examination, but also to highlight that overdocumentation of clinical findings not obtained during patient encounters may jeopardize patient safety. To investigate overall and domain pass/fail patterns of candidates who misrepresented clinical findings with regard to past and subsequent performance on COMLEX-USA Level 2-PE. Specifically, to investigate what percentage of candidates failed because of misrepresentation on first attempts and how they performed on subsequent administrations, as well as the previous performance patterns of candidates who failed because of misrepresentation on examination retakes. Historical records from NBOME's COMLEX-USA Level 2-PE database (testing cycles 2007-2008, 2008-2009, and 2009-2010) were used to analyze overall and domain pass/fail patterns of candidates who failed at least once because of misrepresentation of clinical findings. Of the 24 candidates who failed because of misrepresentation of postencounter (SOAP) notes, 20 candidates (83%) were first-time examinees. Four candidates (17%) were repeating the examination, 2 of whom were making a third attempt to pass. Among these 20 candidates who failed because of misrepresentation of clinical findings on their first attempt, 19 passed on their next attempt. At the time of study analysis, all but 2 candidates eventually passed the examination in subsequent attempts. Among candidates found to have misrepresented clinical findings on postencounter written documentation on COMLEX-USA Level 2-PE, no pattern existed between their past or subsequent performance with regard to overall or domain pass/fail results. The vast

Definitions of fuel poverty: Implications for policy

International Nuclear Information System (INIS)

Moore, Richard

2012-01-01

This paper outlines why the definition of fuel poverty is important in policy formulation and describes how the Government's current definitions evolved from the original concept. It discusses the determination of income and fuel costs and the possibilities for a relative and common European measure. It examines problems inherent in assessing fuel costs as a percentage of income and puts forward the arguments for a ‘budget standard’ approach. The paper illustrates how the size of the problem depends on the definition and chosen threshold and suggests advantages for a rating scale. It illustrates how the income composition and thresholds also govern the distribution of the target populations and the relative importance of the main causal factors, and examines the consequent policy implications. It explores the definition of vulnerable households and the importance of severity and questions whether the UK fuel poverty strategy is targeted at households least able to afford their fuel costs (as the name implies) or primarily those at risk from excess winter and summer mortality and morbidity. Finally, after examining the role of supplementary indicators, it looks at the opportunities for changing the definition and comments on the Government review of the definition and targets. - Highlights: ► There are major failings in the existing official definitions of fuel poverty. ► expressing fuel costs as a percentage of income is a poor indicator of fuel poverty. ► A budget standard approach provides a more consistent, meaningful and fairer measure. ► The scale and nature of the problem changes dramatically with different definitions. ► The definition is crucial to the mix of policies and allocation of resources required.

Log-binomial models: exploring failed convergence.

Science.gov (United States)

Williamson, Tyler; Eliasziw, Misha; Fick, Gordon Hilton

2013-12-13

Relative risk is a summary metric that is commonly used in epidemiological investigations. Increasingly, epidemiologists are using log-binomial models to study the impact of a set of predictor variables on a single binary outcome, as they naturally offer relative risks. However, standard statistical software may report failed convergence when attempting to fit log-binomial models in certain settings. The methods that have been proposed in the literature for dealing with failed convergence use approximate solutions to avoid the issue. This research looks directly at the log-likelihood function for the simplest log-binomial model where failed convergence has been observed, a model with a single linear predictor with three levels. The possible causes of failed convergence are explored and potential solutions are presented for some cases. Among the principal causes is a failure of the fitting algorithm to converge despite the log-likelihood function having a single finite maximum. Despite these limitations, log-binomial models are a viable option for epidemiologists wishing to describe the relationship between a set of predictors and a binary outcome where relative risk is the desired summary measure. Epidemiologists are encouraged to continue to use log-binomial models and advocate for improvements to the fitting algorithms to promote the widespread use of log-binomial models.

SECA Coal-Based Systems - FuelCell Energy, Inc.

Energy Technology Data Exchange (ETDEWEB)

Ayagh, Hossein [Fuelcell Energy, Inc., Danbury, CT (United States)

2014-01-31

The overall goal of this U.S. Department of Energy (DOE) sponsored project is the development of solid oxide fuel cell (SOFC) cell and stack technology suitable for use in highly-efficient, economically-competitive central generation power plant facilities fueled by coal synthesis gas (syngas). This program incorporates the following supporting objectives: • Reduce SOFC-based electrical power generation system cost to $700 or less (2007 dollars) for a greater than 100 MW Integrated Gasification Fuel Cell (IGFC) power plant, exclusive of coal gasification and CO₂ separation subsystem costs. • Achieve an overall IGFC power plant efficiency of at least 50%, from coal (higher heating value or HHV) to AC power (exclusive of CO₂ compression power requirement). • Reduce the release of CO2 to the environment in an IGFC power plant to no more than 10% of the carbon in the syngas. • Increase SOFC stack reliability to achieve a design life of greater than 40,000 hours. At the inception of the project, the efforts were focused on research, design and testing of prototype planar SOFC power generators for stationary applications. FuelCell Energy, Inc. successfully completed the initial stage of the project by meeting the program metrics, culminating in delivery and testing of a 3 kW system at National Energy Technology Laboratory (NETL). Subsequently, the project was re-aligned into a three phase effort with the main goal to develop SOFC technology for application in coal-fueled power plants with >90% carbon capture. Phase I of the Coal-based efforts focused on cell and stack size scale-up with concurrent enhancement of performance, life, cost, and manufacturing characteristics. Also in Phase I, design and analysis of the baseline (greater than 100 MW) power plant system—including concept identification, system definition, and cost analysis—was conducted. Phase II efforts focused on development of a ≥25 kW SOFC stack tower incorporating

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented

A model for fission product distribution in CANDU fuel

International Nuclear Information System (INIS)

Muzumdar, A.P.

1983-01-01

This paper describes a model to estimate the distribution of active fission products among the UO 2 grains, grain-boundaries, and the free void spaces in CANDU fuel elements during normal operation. This distribution is required for the calculation of the potential release of activity from failed fuel sheaths during a loss-of-coolant accident. The activity residing in the free spaces (''free'' inventory) is available for release upon sheath rupture, whereas relatively high fuel temperatures and/or thermal shock are required to release the activity in the grain boundaries or grains. A preliminary comparison of the model with the data from in-reactor sweep-gas experiments performed in Canada yields generally good agreement, with overprediction rather than under prediction of radiologically important isotopes, such as I 131 . The model also appears to generally agree with the ''free'' inventory release calculated using ANS-5.4. (author)

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

Energy Technology Data Exchange (ETDEWEB)

Lanning, D. D.

1981-09-01

This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527.

Households facing constraints. Fuel poverty put into context

Energy Technology Data Exchange (ETDEWEB)

Dubois, Ute [ISG Business School, Paris (France); Meier, Helena [Koeln Univ. (Germany). Energiewirtschaftliches Inst.

2014-02-15

The present paper discusses the concept of fuel poverty taking into account the arbitrages made by households when they are facing economic constraints. Fuel poverty is still lacking a common definition throughout Europe: while the UK and France have (different) official definitions, there is still no definition in a country like Germany, or at the European level. Where definitions exist, they often consider that fuel poor households have high energy needs. The possibility of being fuel poor even without having high energy needs and the various arbitrage possibilities of households - i.e. to under-spend and use too little energy - are not systematically discussed. Our paper tries to fill that gap by putting fuel poverty into the larger context of constraints faced by households. Based on a graphical analysis, it shows that different situations of fuel poverty might occur. It results in the identification of two distinct fuel poverty problems: an ''energy inequality'' problem, reflected by the fact that some households pay disproportionately high energy bills, and an ''energy affordability'' problem that can affect a larger share of the population. It finally explores the two types of fuel poverty for European countries and discusses policy implications.

A study of UO2 wafer fuel for very high-power research reactors

International Nuclear Information System (INIS)

Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

1983-01-01

The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

The failing firm defence: merger policy and entry

OpenAIRE

Mason, Robin; Weeds, Helen

2003-01-01

This Paper considers the 'failing firm defence'. Under this principle, found in most antitrust jurisdictions, a merger that would otherwise be blocked due to its adverse effect on competition is permitted when the firm to be acquired is a failing firm, and an alternative, less detrimental merger is unavailable. Competition authorities have shown considerable reluctance to accept the failing firm defence, and it has been successfully used in just a handful of cases. The Paper considers the def...

Coping Styles of Failing Brunei Vocational Students

Science.gov (United States)

Mundia, Lawrence; Salleh, Sallimah

2017-01-01

Purpose: The purpose of this paper is to determine the prevalence of two types of underachieving students (n = 246) (active failing (AF) and passive failing (PF)) in Brunei vocational and technical education (VTE) institutions and their patterns of coping. Design/methodology/approach: The field survey method was used to directly reach many…

Nuclear Fuels & Materials Spotlight Volume 4

Energy Technology Data Exchange (ETDEWEB)

I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

2014-04-01

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

Shielding considerations for advanced fuel irradiation experiments

International Nuclear Information System (INIS)

Kang, Young-Hwan; Kim, Hee-Moon; Kim, Bong-Goo; Kim, Hark-Rho; Lee, Dong-Soo

2008-01-01

An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 μSv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels. (author)

Oxide fuel fabrication technology development of the FaCT project (5). Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

International Nuclear Information System (INIS)

Ohtsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

2011-01-01

This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production, both of which are key points for the commercialized use of ODS steels as long-life fuel cladding tubes. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed, where the handling of elemental powder is prohibited in the process. (author)

Automatic failure identification of the nuclear power plant pellet fuel

International Nuclear Information System (INIS)

Oliveira, Adriano Fortunato de

2010-01-01

This paper proposed the development of an automatic technique for evaluating defects to help in the stage of fabrication of fuel elements. Was produced an intelligent image analysis for automatic recognition of defects in uranium pellets. Therefore, an Artificial Neural Network (ANN) was trained using segments of histograms of pellets, containing examples of both normal (no fault) and of defectives pellets (with major defects normally found). The images of the pellets were segmented into 11 shares. Histograms were made of these segments and trained the ANN. Besides automating the process, the system was able to obtain this classification accuracy of 98.33%. Although this percentage represents a significant advance ever in the quality control process, the use of more advanced techniques of photography and lighting will reduce it to insignificant levels with low cost. Technologically, the method developed, should it ever be implemented, will add substantial value in terms of process quality control and production outages in relation to domestic manufacturing of nuclear fuel. (author)

Radionuclide release from PWR spent fuel specimens with induced cladding defects

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1984-03-01

Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures

A new disaster victim identification management strategy targeting "near identification-threshold" cases: Experiences from the Boxing Day tsunami.

Science.gov (United States)

Wright, Kirsty; Mundorff, Amy; Chaseling, Janet; Forrest, Alexander; Maguire, Christopher; Crane, Denis I

2015-05-01

The international disaster victim identification (DVI) response to the Boxing Day tsunami, led by the Royal Thai Police in Phuket, Thailand, was one of the largest and most complex in DVI history. Referred to as the Thai Tsunami Victim Identification operation, the group comprised a multi-national, multi-agency, and multi-disciplinary team. The traditional DVI approach proved successful in identifying a large number of victims quickly. However, the team struggled to identify certain victims due to incomplete or poor quality ante-mortem and post-mortem data. In response to these challenges, a new 'near-threshold' DVI management strategy was implemented to target presumptive identifications and improve operational efficiency. The strategy was implemented by the DNA Team, therefore DNA kinship matches that just failed to reach the reporting threshold of 99.9% were prioritized, however the same approach could be taken by targeting, for example, cases with partial fingerprint matches. The presumptive DNA identifications were progressively filtered through the Investigation, Dental and Fingerprint Teams to add additional information necessary to either strengthen or conclusively exclude the identification. Over a five-month period 111 victims from ten countries were identified using this targeted approach. The new identifications comprised 87 adults, 24 children and included 97 Thai locals. New data from the Fingerprint Team established nearly 60% of the total near-threshold identifications and the combined DNA/Physical method was responsible for over 30%. Implementing the new strategy, targeting near-threshold cases, had positive management implications. The process initiated additional ante-mortem information collections, and established a much-needed, distinct "end-point" for unresolved cases. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

[U.S. renewable fuel standard implementation mechanism and market tracking].

Science.gov (United States)

Kang, Liping; Earley, Robert; An, Feng; Zhang, Yu

2013-03-01

U.S. Renewable Fuel Standard (RFS) is a mandatory policy for promoting the utilization of biofuels in road transpiration sector in order to reduce the country's dependency on foreign oil and greenhouse gas emissions. U.S. Environmental Protection Agency (EPA) defines the proportion of renewable fuels according to RFS annual target, and requests obligated parties such like fossil fuel refiner, blenders and importer in the U.S. to complete Renewable Volume Obligation (RVO) every year. Obligated parties prove they have achieved their RVO through a renewable fuels certification system, which generates Renewable Identification Numbers (RINs) for every gallon of qualified renewable fuels produced or imported into U.S., RINs is a key for tracking renewable fuel consumption, which in turn is a key for implementing the RFS in the U.S., separated RINs can be freely traded in market and obligated parties could fulfill their RVO through buying RINs from other stakeholders. This briefing paper highlights RFS policy implementing mechanism and marketing tracking, mainly describes importance of RINs, and the method for generating and tracking RINs by both government and fuels industry participants.

Who Really Failed? Commentary

Science.gov (United States)

Maiuri, Katherine M.; Leon, Raul A.

2012-01-01

Scott Jaschik's (2010) article "Who Really Failed?" details the experience of Dominique Homberger, a tenured faculty member at Louisiana State University (LSU) who was removed from teaching her introductory biology course citing student complaints in regards to "the extreme nature" of the grading policy. This removal has…

Abortion: Strong's counterexamples fail

DEFF Research Database (Denmark)

Di Nucci, Ezio

2009-01-01

This paper shows that the counterexamples proposed by Strong in 2008 in the Journal of Medical Ethics to Marquis's argument against abortion fail. Strong's basic idea is that there are cases--for example, terminally ill patients--where killing an adult human being is prima facie seriously morally...

Fuel failure monitoring system design approach for KALIMER

International Nuclear Information System (INIS)

Song, Soon Ja; Hwang, I. K.; Kwon, Kee Choon

1998-01-01

Fuel Failure Monitoring System (FFMS) detects fission gas and locates failed fuels in Liquid Metal Reactor. This system comprises three subsystems; delayed neutron monitoring, cover gas monitoring, and gas tagging. The purpose of this system is to improve the integrity and availability of the liquid metal plant. In this paper, FFMS was analyzed on detection method and compared with various existing liquid metal plants. Sampling and detecting methods were classified with specific plant types. Several technologies of them was recognized and used in most liquid metal reactors. Detection technology and analysis performance, however, must be improved because of new technology when liquid metal plant is built, but the FFMS design scheme will not be changed. Thereby this paper suggests the design to implement KALIMER(Korea Advanced LIquid MEtal Reactor) FFMS

Energy, equity and the future of the fuel poor

International Nuclear Information System (INIS)

Roberts, Simon

2008-01-01

A warm and adequately-lit home is considered a basic need, together with access to energy-consuming appliances ranging from a fridge to a TV. An underlying tenet of sustainable energy is that such basic needs should be affordably met. Yet low incomes, energy-inefficient housing and appliances and high energy costs mean that roughly 10 per cent of UK households, many of them elderly or with young children, fail to attain this basic standard. These households, which would need to spend more than 10 per cent of their income to attain adequate energy services, are officially defined as 'fuel poor'. Their cold, poorly equipped homes lead to chronic cold-related health conditions, exacerbate social isolation, and may undermine educational achievement. In addition, rural areas have a disproportionately high incidence of fuel poverty. This Review examines the current distribution of energy consumption, its social impacts, and the opportunities to address fuel poverty through improvements to the housing stock. It will then consider potential future developments

AGR fuel pin pellet-clad interaction failure limits and activity release fractions

International Nuclear Information System (INIS)

Hughes, H.; Hargreaves, R.

1985-01-01

The limiting conditions beyond which pellet-clad interaction can flail AGR fuel are described. They have been determined by many experiments involving post-irradiation examination and testing, loop experiments and cycling and up-rating of both individual fuel stringers and the whole WAGR core. The mechanisms causing this interaction are well understood and are quantitatively expressed in computer codes. Strain concentration effects over fuel cracks determine power cycling endurance while additional strain concentrations at clad ridges and from cross pin temperature gradients contribute to up-rating failures. An equation summarising tube burst test data so as to determine the ductility available at any transient is given. The hollow fuel and more ductile clad of the Civil AGR fuel pins leads to a much improved performance over the original fuel design. The Civil AGRs operate well within these limiting conditions and substantial increases beyond the design burn-up are confidently expected. The activity release on pin failure and its development during continued operation of failed fuel have also been investigated. A retention of radioiodine and caesium of 90-99% compared to the noble gases has been demonstrated. Measured fission gas releases into the free volume of Civil AGR fuel pins have been very low (< 0.1%)

Gas tagging system development in Japan

International Nuclear Information System (INIS)

Sekiguchi, N.; Rindo, H.; Akiyama, T.; Miyazawa, T.; Heki, H.

1981-05-01

The Gas tagging method has been considered to be most desirable for a failed fuel location system for the fast breeder reactor, regarding the component reduction in the reactor vessel and rapid location during reactor operation. The gas tagging system has been designed by referring to R and D results obtained in Japan and other countries. The designed system is comprised of tag gas filling pins, cover gas sampling system, tag gas recovery and enrichment system, tag gas analyzer and system control and data handling computers. The main specifications for this system have been decided as follows; 1) Main function is location of failed fuels in core and a part of blanket region, 2) Identification capability is each subassembly, 3) Time for identification is within a few days, 4) Continuous operation with automatic start at fuel failure, 5) Detection sensitivity must cover both gas leak and pin burst. In designing the gas tagging system, the following R and D items were selected; 1) System design study, 2) Tag gas capsule development, 3) Modeling the tag gas behavior in reactor primary cooling system, 4) Tag gas recovery and enrichment system, 5) Computer code development for tag gas isotope ratio change estimation. Details of the Japanese gas tagging system development appear in this paper. (author)

Development of wireless vehicle remote control for fuel lid operation

Science.gov (United States)

Sulaiman, N.; Jadin, M. S.; Najib, M. S.; Mustafa, M.; Azmi, S. N. F.

2018-04-01

Nowadays, the evolution of the vehicle technology had made the vehicle especially car to be equipped with a remote control to control the operation of the locking and unlocking system of the car’s door and rear’s bonnet. However, for the fuel or petrol lid, it merely can be opened from inside the car’s cabin by handling the fuel level inside the car’s cabin to open the fuel lid. The petrol lid can be closed by pushing the lid by hand. Due to the high usage of using fuel lever to open the fuel lid when refilling the fuel, the car driver might encounter the malfunction of fuel lid (fail to open) when pushing or pulling the fuel lever. Thus, the main aim of the research is to enhance the operation of an existing car remote control where the car fuel lid can be controlled using two techniques; remote control-based and smartphone-based. The remote control is constructed using Arduino microcontroller, wireless sensors and XCTU software to set the transmitting and receiving parameters. Meanwhile, the smartphone can control the operation of the fuel lid by communicating with Arduino microcontroller which is attached to the fuel lid using Bluetooth sensor to open the petrol lid. In order to avoid the conflict of instruction between wireless systems with the existing mechanical-based system, the servo motor will be employed to release the fuel lid merely after receiving the instruction from Arduino microcontroller and smartphone. As a conclusion, the prototype of the multipurpose vehicle remote control is successfully invented, constructed and tested. The car fuel lid can be opened either using remote control or smartphone in a sequential manner. Therefore, the outcome of the project can be used to serve as an alternative solution to solve the car fuel lid problem even though the problem rarely occurred.

COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

International Nuclear Information System (INIS)

Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

2002-01-01

1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

Commercializing an alternate transportation fuel: lessons learned from NGV

International Nuclear Information System (INIS)

Flynn, P.C.

2001-01-01

An alternate transportation fuel, compressed natural gas, was adopted in Canada in the mid-1980s due to the unique conditions present at the time. The factors that had an impact on the limited acceptance of the fuel, keeping its rate of adoption below the critical point were examined in this paper. It was revealed that a lack of infrastructure to support converted vehicles was the deciding factor. Existing refueling stations failed to become profitable, preventing further investment in refueling facilities and resulting in depressed sales of converted vehicles. Excessive parts markup by conversion dealers was another major hurdle, as was exaggerated claims for environmental and economic benefits. In addition, promotional programs were poorly designed. In the late 1980s, the relative values of oil and natural gas shifted, lowering the momentum from sales of conversions. The consequence was major players leaving the market and natural gas remained on the fringe in both Canada and the United States. Different alternate transportation fuels, including electricity and hydrogen, are being favored by new technologies and driving forces. The growth to commercial viability for those fuels will likely be influenced by some of the factors that played a role in the fate of natural gas as a transportation fuel. 4 refs., 1 fig

The origin and fate of organic pollutants from the combustion of alternative fuels

International Nuclear Information System (INIS)

1995-06-01

The overall objective of this project is to determine the impact of alternative fuels on air quality, particularly ozone formation. The objective will be met through three steps: (1) qualitative identification of alternative fuel combustion products, (2) quantitative measurement of specific emission levels of these products, and (3) determination of the fate of the combustion products in the atmosphere. The alternative fuels of interest are methanol, ethanol, natural gas, and LP gas. The role of the University of Dayton Research Institute (UDRI) in this project is two-fold. First, fused silica flow reactor instrumentation is being used to obtain both qualitative identification and quantitative data on the thermal degradation products from the fuel-lean (oxidative), stoichiometric, and fuel-rich (pyrolytic) decomposition of methanol, ethanol, liquefied petroleum gas, and natural gas. Secondly, a laser photolysis/laser-induced fluorescence (LP/LIF) apparatus is being used to determine the rates and mechanisms of reaction of selected degradation products under atmospheric conditions. This draft final report contains the results of the second year of the study. The authors initially discuss the results of their flow reactor studies. This is followed by a discussion of the initial results from their LP/LIF studies of the reaction of hydroxyl (OH) radicals with methanol and ethanol. In the coming year, they plan to obtain quantitative data on the oxidation of methyl-t-butyl-ether and reformulated gasoline under fuel-lean, stoichiometric, and fuel-rich conditions. They also plan to conduct a mechanistic analysis of the reaction of OH with acetaldehyde and formaldehyde over an extended temperature range

A collapse mode of failure in powder-filled fuel elements

International Nuclear Information System (INIS)

Feraday, M.A.; Chalder, G.H.

1964-01-01

Two swaged fuel elements containing crushed, fused UO 2 powder were irradiated in a pressurized water loop at high heat ratings (∫Kdθ = 48 w/cm). The fuel elements were 2.0 cm in diameter and were sheathed in nickel-free Zircaloy--2 of 0.038 cm thickness. One element failed when the sheath ruptured at the top of a longitudinal ridge in the sheath after a burn-up of approximately 2550 MWd/TeU. No evidence was found that outgassing of the UO 2 contributed to the failure. Dimensional and structural changes observed in the fuel elements led to the conclusion that ridging of the sheath resulted from the action of coolant pressure on the diametral clearance formed by sintering and shrinkage of the UO 2 . Failure resulted due to severe local deformation accompanying one or more power cycles following ridge formation. (author)

Factors influencing sentinel lymph node identification failure in breast cancer surgery

DEFF Research Database (Denmark)

Straalman, K.; Kristoffersen, U.S.; Galatius, H.

2008-01-01

The purpose of this study was to investigate potential risk factors for failed sentinel lymph node identification in breast cancer surgery. Patient characteristics, tumour characteristics, surgeon experience and detection success/failure were registered at 748 sentinel lymph node biopsy procedures...... at our inpatient clinic. Data were analysed with backward stepwise multiple logistic regression with a cut-off point of p

Data-Driven Photovoltaic System Modeling Based on Nonlinear System Identification

Directory of Open Access Journals (Sweden)

Ayedh Alqahtani

2016-01-01

Full Text Available Solar photovoltaic (PV energy sources are rapidly gaining potential growth and popularity compared to conventional fossil fuel sources. As the merging of PV systems with existing power sources increases, reliable and accurate PV system identification is essential, to address the highly nonlinear change in PV system dynamic and operational characteristics. This paper deals with the identification of a PV system characteristic with a switch-mode power converter. Measured input-output data are collected from a real PV panel to be used for the identification. The data are divided into estimation and validation sets. The identification methodology is discussed. A Hammerstein-Wiener model is identified and selected due to its suitability to best capture the PV system dynamics, and results and discussion are provided to demonstrate the accuracy of the selected model structure.

Development of Coated Particle Fuel Technology

International Nuclear Information System (INIS)

Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

2009-04-01

UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

Comparison of potential radiological impacts of 233U and 239Pu fuel cycles

International Nuclear Information System (INIS)

Meyer, H.R.; Little, C.A.; Witherspoon, J.P.; Till, J.E.

1979-01-01

Nuclear fuel cycles utilizing 233 U are currently the subject of considerable interest in the United States. This paper focuses on the identification of significant differences between the off-site radiological hazards posed by 232 Th/ 233 U (Th/U) and 238 U/ 239 Pu (U/Pu) fuel cycles, and represents a portion of our involvement in the Nonproliferation Alternative Systems Assessment Program (NASAP), to be used in support of the International Fuel Cycle Evaluation (INFCE). The major contributors to radiological dose are likely to be uranium mining and milling (58.5% of total fuel cycle dose), reprocessing (33.9%), and light-water reactor power generation (7.3%). The remainder of the cycle, including enrichment processes, fuel fabrication, transportation, and waste management, contributes only 0.3% to total estimated fuel cycle dose

Issues related to EM management of DOE spent nuclear fuel

International Nuclear Information System (INIS)

Abbott, D.G.; Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M.

1993-07-01

This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE's SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE's national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references

Breaking the Failed-State Cycle

National Research Council Canada - National Science Library

Haims, Marla C; Gompert, David C; Treverton, Gregory F; Stearns, Brooke K

2008-01-01

In their research and field experience, the authors have observed a wide gulf separating the treatment of the security problems of failed states from the treatment of those states economic problems...

Co-Optimization of Fuels & Engines: Fuel Blendstocks with the Potential to Optimize Future Gasoline Engine Performance; Identification of Five Chemical Families for Detailed Evaluation

Energy Technology Data Exchange (ETDEWEB)

Farrell, John T [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Holladay, John [Pacific Northwest National Laboratory; Wagner, Robert [Oak Ridge National Laboratory

2018-04-18

The U.S. Department of Energy's (DOE's) Co-Optimization of Fuels & Engines (Co-Optima) initiative is conducting the early-stage research needed to accelerate the market introduction of advanced fuel and engine technologies. The research includes both spark-ignition (SI) and compression-ignition (CI) combustion approaches, targeting applications that impact the entire on-road fleet (light-, medium-, and heavy-duty vehicles). The initiative's major goals include significant improvements in vehicle fuel economy, lower-cost pathways to reduce emissions, and leveraging diverse U.S. fuel resources. A key objective of Co-Optima's research is to identify new blendstocks that enhance current petroleum blending components, increase blendstock diversity, and provide refiners with increased flexibility to blend fuels with the key properties required to optimize advanced internal combustion engines. This report identifies eight representative blendstocks from five chemical families that have demonstrated the potential to increase boosted SI engine efficiency, meet key fuel quality requirements, and be viable for production at commercial scale by 2025-2030.

Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

International Nuclear Information System (INIS)

Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

2005-01-01

From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented

System of leak inspection of irradiated fuel; Sistema de inspeccion de fuga de combustible irradiado

Energy Technology Data Exchange (ETDEWEB)

Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: rmr@nuclear.inin.mx

2007-07-01

The International Atomic Energy Agency (IAEA) through the project RLA/04/18 Irradiated Fuel Management in Research reactors, recommended among other that the participant countries (Brazil, Argentina, Chile, Peru and Mexico), develop the sipping tool to generate registrations of the state that keep the irradiated fuels in the facilities of each country. The TRIGA Mark lll Reactor (RTMIII) Department, generated a project that it is based on the dimensions of the used fuel by the RTMIII, for design and to build an inspection system of irradiated fuel well known as SIPPING. This technique, provides a high grade of accuracy in the detection of gassy fission products or liquids that escape from the enveloping of fuels that have flaws or flights. The operation process of the SIPPING is carried out generating the migration of fission products through the creation of a pressure differential gas or vacuum to identify fuel assemblies failed by means of the detection of the xenon and/or krypton presence. The SIPPING system, is a device in revolver form with 4 tangential nozzles, which will discharge the fluid between the external surface of the enveloping of the fuel and the interior surface of the encircling one; the device was designed with independent pieces, with threaded joining and with stamps to impede flights of the fluid toward the exterior of the system. The System homogenizes and it distributes the fluid pressure so that the 4 nozzles work to equality of conditions, for what the device was designed in 3 pieces, an internal that is denominated revolver, one external that calls cover, and a joining called mamelon that will unite with the main encircling of the system. The detection of fission products in failed fuels, its require that inside the encircling one where the irradiated fuel element is introduced, be generated a pressure differential of gas or vacuum, and that it allows the samples extraction of water. For what generated a top for the encircling with the

Control of helium activity in the fuel reactor channels; Kontrola aktivnosti heliuma u tehnoloskim kanalima

Energy Technology Data Exchange (ETDEWEB)

Vidmar, M; Milosevic, M; Hadzic, S [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

1961-02-15

The objective of this task was to study the possibility of detecting a damaged fuel channel, and to introduce automated procedure for continuous control of reactor channels during operation. The existing control systems at the RA reactor (permanent control of heavy water and helium activity, radiation monitoring of heavy water and helium system, measurements of fire damp gas percent) are not sufficient for fast detection of fuel element failures. Since a 'hot' fuel channel cannot be removed from the core because it should be cooled in the core by heavy water circulation, it is not possible to prevent contamination of heavy water by fission products. It is concluded that it is not indispensable to detect the failed fuel element promptly, i.e. that tome is not a critical issue.

Modeling of a HTPEM fuel cell using Adaptive Neuro-Fuzzy Inference Systems

DEFF Research Database (Denmark)

Justesen, Kristian Kjær; Andreasen, Søren Juhl; Sahlin, Simon Lennart

2015-01-01

In this work an Adaptive Neuro-Fuzzy Inference System (ANFIS) model of the voltage of a fuel cell is developed. The inputs of this model are the fuel cell temperature, current density and the carbon monoxide concentration of the anode supply gas. First an identification experiment which spans...... the expected operating range of the fuel cell is performed in a test station. The data from this experiment is then used to train ANFIS models with 2, 3, 4 and 5 membership functions. The performance of these models is then compared and it is found that using 3 membership functions provides the best compromise...

Laparoscopic revision of failed antireflux operations.

Science.gov (United States)

Serafini, F M; Bloomston, M; Zervos, E; Muench, J; Albrink, M H; Murr, M; Rosemurgy, A S

2001-01-01

A small number of patients fail fundoplication and require reoperation. Laparoscopic techniques have been applied to reoperative fundoplications. We reviewed our experience with reoperative laparoscopic fundoplication. Reoperative laparoscopic fundoplication was undertaken in 28 patients, 19 F and 9 M, of mean age 56 years +/- 12. Previous antireflux procedures included 19 open and 12 laparoscopic antireflux operations. Symptoms were heartburn (90%), dysphagia (35%), and atypical symptoms (30%%). The mean interval from antireflux procedure to revision was 13 months +/- 4.2. The mean DeMeester score was 78+/-32 (normal 14.7). Eighteen patients (64%) had hiatal breakdown, 17 (60%) had wrap failure, 2 (7%) had slipped Nissen, 3 (11%) had paraesophageal hernias, and 1 (3%) had an excessively tight wrap. Twenty-five revisions were completed laparoscopically, while 3 patients required conversion to the open technique. Complications occurred in 9 of 17 (53%) patients failing previous open fundoplications and in 4 of 12 patients (33%) failing previous laparoscopic fundoplications and included 15 gastrotomies and 1 esophagotomy, all repaired laparoscopically, 3 postoperative gastric leaks, and 4 pneumothoraces requiring tube thoracostomy. No deaths occurred. Median length of stay was 5 days (range 2-90 days). At a mean follow-up of 20 months +/- 17, 2 patients (7%) have failed revision of their fundoplications, with the rest of the patients being essentially asymptomatic (93%). The results achieved with reoperative laparoscopic fundoplication are similar to those of primary laparoscopic fundoplications. Laparoscopic reoperations, particularly of primary open fundoplication, can be technically challenging and fraught with complications. Copyright 2001 Academic Press.

Effects of consumer subsidy on household fuel switching from coal to cleaner fuels: A case study for anthracites in Korea

International Nuclear Information System (INIS)

Park, Hojeong; Kwon, Hyucksoo

2011-01-01

The Korean coal industry is in a transition under low carbon policy through the steady reduction of coal production. Since consumer subsidy for the consumption of anthracites briquette in low-income households caused a distortion in domestic coal market, the so-called coupon program will be the first target in energy reform policy in order to induce fuel switching from anthracites to alternative clean energy. This paper tries to identify various factors that influence households' fuel switching decision. Disutility from briquette consumption is also considered as an important factor. Using the 2007 census data on briquette-consuming households, it is found that the coupon program provides an adverse effect to switching fuels to clean energy while the disutility of briquettes is positively associated with the probability of fuel switching. However, the empirical finding suggests that the policy alone attempting to remove coupon program may fail to switch fuels unless the cost of boiler changes is substantially reduced through the provision of accessible networks to alternative energy sources. It indicates that reform policy for consumer subsidy must be understood in line with more comprehensive regional energy plans to resolve energy poverty issue. - Research highlights: → Various factors are identified for low income households' fuel switching from coal to clean energy. → Coupon program and accessibility to alternative energy sources are considered; Result shows that accessible network to alternative clean energy sources is essential for low income group. → More comprehensive regional energy plans are required to resolve energy poverty issue.

Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

International Nuclear Information System (INIS)

1990-03-01

This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

Experience of development of the methods and equipment and the prospects for creation of WWER fuel examination stands

International Nuclear Information System (INIS)

Pavlov, S.; Smirnov, V.

1998-01-01

The report presents the basic methods and equipment developed for inspection of the fuel elements and fuel assemblies in the spent fuel pools. It considers their characteristics and results of the tests under laboratory and experimental fuel examination stand conditions. In particular, the following techniques are presented: visual inspection, measurement of the geometrical dimensions, definition of the form change in fuel assemblies and fuel elements, detection of the failed fuel elements, etc. The experience of the experimental fuel examination stand operation is generalized. The concept of the creation of the WWER-440 and WWER-1000 FA and FE inspection stands is presented. The concept is based on the modular principle which runs as follows. A set of the basic functional blocks is being developed based on which it is possible to make such a stand configuration which is necessary to fulfil the specific program of the examination at the particular nuclear power plant. (author)

Technology readiness levels for advanced nuclear fuels and materials development

Energy Technology Data Exchange (ETDEWEB)

Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

2017-03-15

Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

Rescue of failed filtering blebs with ab interno trephination.

Science.gov (United States)

Shihadeh, Wisam A; Ritch, Robert; Liebmann, Jeffrey M

2006-06-01

We evaluated the effectiveness of ab interno automated trephination as a technique for rescuing failed mature filtering blebs. A retrospective chart review of 40 failed blebs of 38 patients who had a posttrephination follow-up period of at least 3 months was done. With success defined as intraocular pressure (IOP) control with other modalities of management. Complications were few. We believe that ab interno trephination is an excellent option for rescuing selected failed filtering blebs.

77 FR 9846 - Source of Income From Qualified Fails Charges

Science.gov (United States)

2012-02-21

... temporary regulations noted that no trading practice existed at that time for fails charges on securities other than Treasuries, but that if a fails charge trading practice pertaining to other securities was... sources within the United States, and the income from the qualified fails charge is treated as effectively...

One- and two-dimension effects on fuel pin lifetime

International Nuclear Information System (INIS)

Stephen, J.D.; Biancheria, A.; Leibnitz, D.; O'Reilly, B.D.; Liu, Y.Y.; Labar, M.P.; Gneiting, B.C.

1979-01-01

Lifetime, or breach of the cladding, is a difficult performance limit to establish in fuel pin design. The significant benefits of high plant capacity factor favor conservative design to eliminate downtime or partial power operation caused by the breach limit; however, overly conservative design produces significant penalties. The LIFE system is being applied to help understand the range between operation and breach so that appropriate design margins can be selected. Standards are being developed in the USA to assure the structural integrity of all core components. These standards will provide guidelines to account for the failure mechanisms observed in the high temperature, high fluence core environment. The work to date indicates that creep rupture is the most important failure mechanism for mixed-oxide fuel pins during normal operation and slow power changes. The local cumulative creep rupture damage fraction (CDF) has been adopted as the parameter to assess the approach to failure. Several oxide breached pins and siblings have been studied For example, the P23B-73 pin was an FFTR driver design pin irradiated in EBR-II which failed at 10 at,% burnup. Initial evaluation based on LIFE3 led to the conclusion that the pin should not have failed. Further analyses determined the sensitivity of the breach prediction to the time-to-rupture correlation, cladding temperature, and fuel-fission product swelling (which had not been modeled in LIFE3). The uncertainties in the time-to-rupture correlation have been established. But LIFE is a one-dimensional model. The TWOD code is complete, and development of the best way to couple LIFE and TWOD for lifetime analysis is in progress. Two preliminary conclusions from analysis of representative oxide pin geometries are, first, that the circumferential stress distribution may not peak at the hot spot, but the damage (CDF) does. And second, that the effect of stress concentrations near fuel cracks on cladding creep damage is small

The Statistical Analysis Techniques to Support the NGNP Fuel Performance Experiments

International Nuclear Information System (INIS)

Pham, Bihn T.; Einerson, Jeffrey J.

2010-01-01

This paper describes the development and application of statistical analysis techniques to support the AGR experimental program on NGNP fuel performance. The experiments conducted in the Idaho National Laboratory's Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel/graphite temperature) is regulated by the He-Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the SAS-based NGNP Data Management and Analysis System (NDMAS) for automated processing and qualification of the AGR measured data. The NDMAS also stores daily neutronic (power) and thermal (heat transfer) code simulation results along with the measurement data, allowing for their combined use and comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the target quantity (fuel temperature) within a given range.

The statistical analysis techniques to support the NGNP fuel performance experiments

Energy Technology Data Exchange (ETDEWEB)

Pham, Binh T., E-mail: Binh.Pham@inl.gov; Einerson, Jeffrey J.

2013-10-15

This paper describes the development and application of statistical analysis techniques to support the Advanced Gas Reactor (AGR) experimental program on Next Generation Nuclear Plant (NGNP) fuel performance. The experiments conducted in the Idaho National Laboratory’s Advanced Test Reactor employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule. The tests are instrumented with thermocouples embedded in graphite blocks and the target quantity (fuel temperature) is regulated by the He–Ne gas mixture that fills the gap volume. Three techniques for statistical analysis, namely control charting, correlation analysis, and regression analysis, are implemented in the NGNP Data Management and Analysis System for automated processing and qualification of the AGR measured data. The neutronic and thermal code simulation results are used for comparative scrutiny. The ultimate objective of this work includes (a) a multi-faceted system for data monitoring and data accuracy testing, (b) identification of possible modes of diagnostics deterioration and changes in experimental conditions, (c) qualification of data for use in code validation, and (d) identification and use of data trends to support effective control of test conditions with respect to the test target. Analysis results and examples given in the paper show the three statistical analysis techniques providing a complementary capability to warn of thermocouple failures. It also suggests that the regression analysis models relating calculated fuel temperatures and thermocouple readings can enable online regulation of experimental parameters (i.e. gas mixture content), to effectively maintain the fuel temperature within a given range.

Why Companies Fail? The Boiling Frog Syndrome

OpenAIRE

Ozcan, Rasim

2018-01-01

Why nations fail? An answer is given by Acemoglu and Robinson (2012) by pointing out the importance of institutions for an economy that leads to innovations for economic growth. Christensen (2012) asks a similar question for a firm and diagnoses why companies fail. In this study, I relate Acemoglu and Robinson (2012) with Christensen (2012) in order to better understand how to make companies more prosperous, more powerful, healthier, and live longer via innovations.

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

Richins, W.D.; Miller, G.K.

1995-12-01

This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

Towards a lumped reaction model for future designer fuels

Energy Technology Data Exchange (ETDEWEB)

Vandersickel, A.; Wright, Y.M.; Boulouchos, K. [ETH Zurich, Aerothermochemistry and Combustion Systems Laboratory, Zurich (Switzerland). Inst. of Energy Technology

2009-07-01

The homogeneous charge compression ignition (HCCI) is one of the most promising engine processes to simultaneously reduce nitrogen oxide and soot emissions. However, its applicability is hindered by its relatively limited operating range. Designer fuels offer unique possibilities for tailoring evaporation and auto-ignition properties, offering a means to control and expand the HCCI operation range. The identification of HCCI relevant fuel properties as well as the definition of a new fuel index able to describe a fuels suitability for HCCI was required in order to develop such designer fuels. This paper discussed a numerical and experimental investigation of a large set of technical fuels covering a wide range of properties. The paper discussed mechanism development approaches, optimization of the lumped mechanism, and and results. Zheng's 7-step reaction mechanism was successfully coupled with a genetic optimization algorithm and fitted to n-heptane ignition delay data. It was concluded that the presented coupled approach could improve the predictive quality of the model and demonstrate that the Zheng model was sufficiently elaborate to emulate the influence of temperature, pressure, exhaust gas recirculation and lambda on ignition. 8 refs., 1 tab., 3 figs.

Towards a lumped reaction model for future designer fuels

International Nuclear Information System (INIS)

Vandersickel, A.; Wright, Y.M.; Boulouchos, K.

2009-01-01

The homogeneous charge compression ignition (HCCI) is one of the most promising engine processes to simultaneously reduce nitrogen oxide and soot emissions. However, its applicability is hindered by its relatively limited operating range. Designer fuels offer unique possibilities for tailoring evaporation and auto-ignition properties, offering a means to control and expand the HCCI operation range. The identification of HCCI relevant fuel properties as well as the definition of a new fuel index able to describe a fuels suitability for HCCI was required in order to develop such designer fuels. This paper discussed a numerical and experimental investigation of a large set of technical fuels covering a wide range of properties. The paper discussed mechanism development approaches, optimization of the lumped mechanism, and and results. Zheng's 7-step reaction mechanism was successfully coupled with a genetic optimization algorithm and fitted to n-heptane ignition delay data. It was concluded that the presented coupled approach could improve the predictive quality of the model and demonstrate that the Zheng model was sufficiently elaborate to emulate the influence of temperature, pressure, exhaust gas recirculation and lambda on ignition. 8 refs., 1 tab., 3 figs.

PEM fuel cell model suitable for energy optimization purposes

International Nuclear Information System (INIS)

Caux, S.; Hankache, W.; Fadel, M.; Hissel, D.

2010-01-01

Many fuel cell stack models or fuel cell system models exist. A model must be built with a main objective, sometimes for accurate electro-chemical behavior description, sometimes for optimization procedure at a system level. In this paper, based on the fundamental reactions present in a fuel cell stack, an accurate model and identification procedure is presented for future energy management in a Hybrid Electrical Vehicle (HEV). The proposed approach extracts all important state variables in such a system and based on the control of the fuel cell's gas flows and temperature, simplification arises to a simple electrical model. Assumptions verified due to the control of the stack allow simplifying the relationships within keeping accuracy in the description of a global fuel cell stack behavior from current demand to voltage. Modeled voltage and current dynamic behaviors are compared with actual measurements. The obtained accuracy is sufficient and less time-consuming (versus other previously published system-oriented models) leading to a suitable model for optimization iterative off-line algorithms.

PEM fuel cell model suitable for energy optimization purposes

Energy Technology Data Exchange (ETDEWEB)

Caux, S.; Hankache, W.; Fadel, M. [LAPLACE/CODIASE: UMR CNRS 5213, Universite de Toulouse - INPT, UPS, - ENSEEIHT: 2 rue Camichel BP7122, 31071 Toulouse (France); CNRS, LAPLACE, F-31071 Toulouse (France); Hissel, D. [FEMTO-ST ENISYS/FCLAB, UMR CNRS 6174, University of Franche-Comte, Rue Thierry Mieg, 90010 Belfort (France)

2010-02-15

Many fuel cell stack models or fuel cell system models exist. A model must be built with a main objective, sometimes for accurate electro-chemical behavior description, sometimes for optimization procedure at a system level. In this paper, based on the fundamental reactions present in a fuel cell stack, an accurate model and identification procedure is presented for future energy management in a Hybrid Electrical Vehicle (HEV). The proposed approach extracts all important state variables in such a system and based on the control of the fuel cell's gas flows and temperature, simplification arises to a simple electrical model. Assumptions verified due to the control of the stack allow simplifying the relationships within keeping accuracy in the description of a global fuel cell stack behavior from current demand to voltage. Modeled voltage and current dynamic behaviors are compared with actual measurements. The obtained accuracy is sufficient and less time-consuming (versus other previously published system-oriented models) leading to a suitable model for optimization iterative off-line algorithms. (author)

Molten Fuel Mass Assessment for Channel Flow Blockage Event in CANDU6

International Nuclear Information System (INIS)

Lee, Kwang Ho; Kim, Yong Bae; Choi, Hoon; Park, Dong Hwan

2011-01-01

In CANDU6, a fuel channel flow blockage causes a sudden reduction of flow through the blocked channel. Depending on the severity of the blockage, the reduced flow through the channel can result in severe heat up of the fuel, hence possibly leading to pressure tube and calandria tube failure. If the calandria tube does not fail the fuel and sheath would continue to heat up, and ultimately melting could occur. Eventually, molten material runs down onto the pressure tube. Even a thin layer of molten material in contact with the pressure tube causes the pressure tube and calandreia tube to heat up rapidly. The thermal transient is so rapid that failure temperatures are reached quickly. After channel failure, the contents of the channel, consisting of superheated coolant, fission products and possibly overheated of molten fuel, are rapidly discharged into the moderator. Fuel discharged into the moderator is quenched and cooled. The rapid discharge of hot fuel and coolant into the calandria causes the moderator pressure and temperature to increase, which may cause damage to some in-core components. Thus, the assessment results of molten fuel mass are inputs to the in-core damage analysis. In this paper, the analysis methodology and results of molten fuel mass assessment for the channel flow blockage event are presented

Current state of knowledge in radiolysis effects on spent fuel corrosion

International Nuclear Information System (INIS)

Christensen, H.; Sunder, S.

1998-09-01

Literature data on the effect of water radiolysis products on spent fuel oxidation and dissolution have been reviewed. Effects of γ-radiolysis, α-radiolysis and dissolved O 2 or H 2 O 2 in unirradiated solutions have been discussed separately. Also the effect of carbonate in γ-irradiated solutions and radiolysis effects on leaching of spent fuels have been reviewed. In addition a radiolysis model for calculation of corrosion rates of UO 2 , presented previously, has been discussed. The model has been shown to give a good agreement between calculated and measured corrosion rates in the case of γ-radiolysis and in unirradiated solutions of dissolved oxygen or hydrogen peroxide. The model has failed to predict the results of α-radiolysis. In a recent study it was shown that the model gave a good agreement with measured corrosion rates of spent fuel exposed in deionized water

Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

International Nuclear Information System (INIS)

Olander, D.R.

1982-01-01

In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

Identification of nuclear plant temperatures. Feedback parameters using experimental data

International Nuclear Information System (INIS)

Abdel Hamid, Sayed.

1981-09-01

This work is concerned with the identification of the fuel and moderator reactivity feedback coefficients of a Pressurized Water Reactor (PWR) using actual measurements. The main aim of this study is to examine the possibility to use a simplified model representing the reactor dynamics, which can be simulated on minicomputer and to supply an identification algorithm to get the feedback coefficients of PWR. The theoretical model of a PWR is built from the space independent reactor kinetics equation associated with six delayed neutron groups equations, as well as twelve equations describe the heat balance for the fuel and the moderator inside the reactor core, assuming that the core is composed from six successive axial zones. The reactor is externally perturbed by moving its control rods, and the corresponding changes in power and temperatures are recorded. The mathematical model has been solved numerically using fifth order Runge-Kutta integration technique by special developed package using Solar 16-40 computer (64 K memory size). As an identification algorithm, the Nelder-Mead Simplex method has been used to minimize the sum of the squares of the differences between measured and calculated reactor power. Hence, the feedback coefficients have been identified from off-line calculations

Economics of Direct Hydrogen Polymer Electrolyte Membrane Fuel Cell Systems

Energy Technology Data Exchange (ETDEWEB)

Mahadevan, Kathyayani

2011-10-04

Battelle's Economic Analysis of PEM Fuel Cell Systems project was initiated in 2003 to evaluate the technology and markets that are near-term and potentially could support the transition to fuel cells in automotive markets. The objective of Battelle?s project was to assist the DOE in developing fuel cell systems for pre-automotive applications by analyzing the technical, economic, and market drivers of direct hydrogen PEM fuel cell adoption. The project was executed over a 6-year period (2003 to 2010) and a variety of analyses were completed in that period. The analyses presented in the final report include: Commercialization scenarios for stationary generation through 2015 (2004); Stakeholder feedback on technology status and performance status of fuel cell systems (2004); Development of manufacturing costs of stationary PEM fuel cell systems for backup power markets (2004); Identification of near-term and mid-term markets for PEM fuel cells (2006); Development of the value proposition and market opportunity of PEM fuel cells in near-term markets by assessing the lifecycle cost of PEM fuel cells as compared to conventional alternatives used in the marketplace and modeling market penetration (2006); Development of the value proposition of PEM fuel cells in government markets (2007); Development of the value proposition and opportunity for large fuel cell system application at data centers and wastewater treatment plants (2008); Update of the manufacturing costs of PEM fuel cells for backup power applications (2009).

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

International Nuclear Information System (INIS)

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO 2 pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel

7 CFR 983.152 - Failed lots/rework procedure.

Science.gov (United States)

2010-01-01

... 7 Agriculture 8 2010-01-01 2010-01-01 false Failed lots/rework procedure. 983.152 Section 983.152..., ARIZONA, AND NEW MEXICO Rules and Regulations § 983.152 Failed lots/rework procedure. (a) Inshell rework procedure for aflatoxin. If inshell rework is selected as a remedy to meet the aflatoxin regulations of this...

Current state of knowledge of water radiolysis effects on spent nuclear fuel corrosion

International Nuclear Information System (INIS)

Christensen, H.; Sunder, S.

2000-07-01

Literature data on the effect of water radiolysis products on spent-fuel oxidation and dissolution are reviewed. Effects of gamma radiolysis, alpha radiolysis, and dissolved O 2 or H 2 O 2 in unirradiated solutions are discussed separately. Also, the effect of carbonate in gamma-irradiated solutions and radiolysis effects on leaching of spent fuel are reviewed. In addition, a kinetic model for calculating the corrosion rates of UO 2 in solutions undergoing radiolysis is discussed. The model gives good agreement between calculated and measured corrosion rates in the case of gamma radiolysis and in unirradiated solutions containing dissolved oxygen or hydrogen peroxide. However, the model fails to predict the results of alpha radiolysis. In a recent study , it was shown that the model gave good agreement with measured corrosion rates of spent fuel exposed in deionized water. The applications of radiolysis studies for geologic disposal of used nuclear fuel are discussed. (author)

Patient identification errors are common in a simulated setting.

Science.gov (United States)

Henneman, Philip L; Fisher, Donald L; Henneman, Elizabeth A; Pham, Tuan A; Campbell, Megan M; Nathanson, Brian H

2010-06-01

We evaluate the frequency and accuracy of health care workers verifying patient identity before performing common tasks. The study included prospective, simulated patient scenarios with an eye-tracking device that showed where the health care workers looked. Simulations involved nurses administering an intravenous medication, technicians labeling a blood specimen, and clerks applying an identity band. Participants were asked to perform their assigned task on 3 simulated patients, and the third patient had a different date of birth and medical record number than the identity information on the artifact label specific to the health care workers' task. Health care workers were unaware that the focus of the study was patient identity. Sixty-one emergency health care workers participated--28 nurses, 16 technicians, and 17 emergency service associates--in 183 patient scenarios. Sixty-one percent of health care workers (37/61) caught the identity error (61% nurses, 94% technicians, 29% emergency service associates). Thirty-nine percent of health care workers (24/61) performed their assigned task on the wrong patient (39% nurses, 6% technicians, 71% emergency service associates). Eye-tracking data were available for 73% of the patient scenarios (133/183). Seventy-four percent of health care workers (74/100) failed to match the patient to the identity band (87% nurses, 49% technicians). Twenty-seven percent of health care workers (36/133) failed to match the artifact to the patient or the identity band before performing their task (33% nurses, 9% technicians, 33% emergency service associates). Fifteen percent (5/33) of health care workers who completed the steps to verify patient identity on the patient with the identification error still failed to recognize the error. Wide variation exists among health care workers verifying patient identity before performing everyday tasks. Education, process changes, and technology are needed to improve the frequency and accuracy of

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

International Nuclear Information System (INIS)

Miron, Adrian; Valentine, Joshua; Christenson, John; Hawwari, Majd; Bhatt, Santosh; Dunzik-Gougar, Mary Lou; Lineberry, Michael

2009-01-01

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), University of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

Energy Technology Data Exchange (ETDEWEB)

Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

2009-10-01

The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1996-01-01

This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process

Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

significantly lower enthalpies than those required for melting, when the fuel burnup exceeds approximately 40 MWd/kgU. This issue is investigated by reviewing all high-burnup UO 2 fuel rods that have failed in RIA simulation tests in the Japanese Nuclear Safety Research Reactor and the French CABRI pulse reactor to date. Data from thirteen failed rods, with burnups between 44 and 64 MWd/kgU, indicate that clad tube failure does not necessarily lead to fuel pellet dispersal. In fact, the data suggest that a peak fuel radial average enthalpy of at least 500 J/gUO 2 is required to expel a significant part (>10 %) of the fuel inventory into the coolant. However, this empirical enthalpy threshold for fuel dispersal from high-burnup fuel rods cannot be directly applied to light water reactors, since the power pulses and/or the cooling conditions used in the pulse tests differ notably from those expected in LWRs under RIA

Identification procedures as a part of death investigation in Turkey.

Science.gov (United States)

Uzün, Ibrahim; Daregenli, Oner; Sirin, Gözde; Müslümanoğlu, Omer

2012-03-01

Forensic identification techniques include the examination of ID cards, the decedent's private belongings, fingerprints, footprints, lip marks, dental findings, red blood cell enzymes, performing photograph matching, facial reconstruction, visual identification, and DNA "fingerprinting." As part of forensic examinations, the identification of corpses that are fresh, decomposed, fragmented, or skeletonized as well as individual body parts and human remains can be requested. Identification becomes a challenging task for forensic terms particularly in mass-disaster situations. Each identification case should be considered to its own merit and the way to do that should be based on the effectiveness and cost of each method used. In Turkey, one of the major duties of the medicolegal system on the investigation of deaths is to identify the deceased if unknown.This study is undertaken to investigate the procedures, as well as their validities, used to deal with individualization of dismembered bodies directly sent to the Council of Forensic Medicine, Ministry of Justice, for autopsy and/or visual identification, as well as those received from peripheral districts for forensic identification. According to the Turkish Penal Procedural Law, a positive identification of the deceased is mandatory before performing an autopsy. According to the law, the ID cards are not taken to be sufficient for recognition of the deceased, and the major way of identification in daily practice is visual identification by a relative or any recognizant person to approve the identification to the prosecutor. If visual identification fails, fingerprints, dental x-rays or body x-rays, and DNA "fingerprinting" can be used to establish identity when compared with known records of the individual obtained by law enforcement.This retrospective study was carried out into 421 dismembered bodies, among 3063 autopsies performed in year 2002 by the Department of Morgue at the Council of Forensic Medicine, with

Programs for Children with Specific Learning Disabilities. P.L. 91-230, Title VI-G. Final Report on Products and Results of Year's Work, July 1, 1973 to June 30, 1974. Fail Save Continuum.

Science.gov (United States)

1974

Presented is the final report of the Learning Disability Fail-Save Continuum Project, designed to provide identification, assessment, and placement services to 378 elementary and secondary level students. Outlined in Section I are the following project components: major revisions from the original proposal (such as implementation of the itinerant…

Activity release during the dry storage of fuel assemblies

International Nuclear Information System (INIS)

Valentine, M.K.; Fettel, W.; Gunther, H.

1991-01-01

This paper reports that wet storage is the predominant storage method in the USA for spent fuel assemblies. Nevertheless, most utilities have stretched their storage capacities and several reactors will lose their full-core reserve in the 90's. A great variety of out-of-pool storage methods already exist, including the FUELSTOR vault-type dry storage concept. A FUELSTOR vault relies on double containment of the spent fuel (intact cladding as the primary containment and sealing of assemblies in canisters filled with an inert gas as the secondary containment) to reduce radiation levels at the outside wall of the vault to less than site boundary levels. Investigation of accident scenarios reveals that radiation release limits are only exceeded following complete failure of all canisters and simultaneous cladding breach for more than 40% of the rods (or for more than 1% of failed rods if massive fuel oxidation occurs following cladding failure). Such failures are considered highly improbable. Thus, it can be concluded that this type of dry storage is safe and individual canister monitoring is not required in the facility

Failing by design.

Science.gov (United States)

McGrath, Rita Gunther

2011-04-01

It's hardly news that business leaders work in increasingly uncertain environments, where failures are bound to be more common than successes. Yet if you ask executives how well, on a scale of one to 10, their organizations learn from failure, you'll often get a sheepish "Two-or maybe three" in response. Such organizations are missing a big opportunity: Failure may be inevitable but, if managed well, can be very useful. A certain amount of failure can help you keep your options open, find out what doesn't work, create the conditions to attract resources and attention, make room for new leaders, and develop intuition and skill. The key to reaping these benefits is to foster "intelligent failure" throughout your organization. McGrath describes several principles that can help you put intelligent failure to work. You should decide what success and failure would look like before you start a project. Document your initial assumptions, test and revise them as you go, and convert them into knowledge. Fail fast-the longer something takes, the less you'll learn-and fail cheaply, to contain your downside risk. Limit the number of uncertainties in new projects, and build a culture that tolerates, and sometimes even celebrates, failure. Finally, codify and share what you learn. These principles won't give you a means of avoiding all failures down the road-that's simply not realistic. They will help you use small losses to attain bigger wins over time.

Nuclear Fuel Cycle System Analysis (II)

Energy Technology Data Exchange (ETDEWEB)

Ko, Won Il; Kwon, Eun Ha; Yoon, Ji Sup; Park, Seong Won

2007-04-15

As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 2nd Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance, economics and technologies. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects of sustainability, environment-friendliness, proliferation-resistance, economics, technologies and so on, which makes the comparison of the options extremely complicated. Taking this into consideration, the paper analyzes all the four fuel cycle options using the Multi-Attribute Utility Theory (MAUT) and the Analytic Hierarchy Process (AHP), methods of Multi-Attribute Decision Making (MADM), that support systematical evaluation of the cases with multi- goals or criteria and that such goals are incompatible with each other. The analysis shows that the GEN-IV Recycle appears to be most competitive.

Weeded Out? Gendered Responses to Failing Calculus.

Science.gov (United States)

Sanabria, Tanya; Penner, Andrew

2017-06-01

Although women graduate from college at higher rates than men, they remain underrepresented in science, technology, engineering, and mathematics (STEM) fields. This study examines whether women react to failing a STEM weed-out course by switching to a non-STEM major and graduating with a bachelor's degree in a non-STEM field. While competitive courses designed to weed out potential STEM majors are often invoked in discussions around why students exit the STEM pipeline, relatively little is known about how women and men react to failing these courses. We use detailed individual-level data from the National Educational Longitudinal Study (NELS) Postsecondary Transcript Study (PETS): 1988-2000 to show that women who failed an introductory calculus course are substantially less likely to earn a bachelor's degree in STEM. In doing so, we provide evidence that weed-out course failure might help us to better understand why women are less likely to earn degrees.

A worldwide fuel strategy by AREVA

International Nuclear Information System (INIS)

Bordy, Michel

2004-01-01

Operating as a global company, inside AREVA the Fuel Sector implements a common strategy among three Business Units of fuel activities. These Business Units which are in Framatome ANP Zirconium, Manufacturing and Design and Sales Units, are operated in Germany (former Siemens activity), in USA (former BWFC Babcock and Wilcox Fuel Co,. and SPC Siemens Power Co. activities), in Belgium and in France (former Framatome activity). They have resources and facilities which are cooperatively working on R and D, engineering, project management, sales and services to achieve synergy on a cross-business basis. Based on its experience of worldwide activities and taking advantage of its diversified fuel design knowledge, Framatome ANP proposes a full range of fuel products and services on the BWR and PWR markets. With the ability to supply all fuel assembly arrays and fuel pellet types, supplemented by the range of stationary and movable core components, and completed by a full-range of on-site fuel services and performance of fuel packing and delivery, Framatome ANP is positioned as a major participant on the world fuel market. Today, Framatome ANP takes advantage of the cross-fertilization in the short term of existing products which include four original PWR fuel designs of HTP TM alloy as the reference material for cladding tubes, guide thimbles, and grids, -- Gradual incorporation of the valuable high-stiffiness MONOBLOC tM guide thimble, -- Progressive integressive integration of the High Mechanical Performance (HMP) Inconel end grid, -- Planned standardization of mechanical components such as nozzles, holddown systems and top and bottom connections. As a continuation of its existing technology, Framatome ANP is developing improved technical features within the scope of the Alliance fuel assembly qualification program. With an irradiation program ranging up to a burnup of 70 MWd/kgU expected to be reached in 2006, Alliance shows excellent behaviour with very low corrosion

DECOFF Probabilities of Failed Operations

DEFF Research Database (Denmark)

Gintautas, Tomas

2015-01-01

A statistical procedure of estimation of Probabilities of Failed Operations is described and exemplified using ECMWF weather forecasts and SIMO output from Rotor Lift test case models. Also safety factor influence is investigated. DECOFF statistical method is benchmarked against standard Alpha-factor...

Postirradiation examination and evaluation of Peach Bottom fuel test element FTE-6

International Nuclear Information System (INIS)

Wallroth, C.F.; Holzgraf, J.F.; Jensen, D.D.

1977-09-01

Fuel test element FTE-6 was irradiated in the Peach Bottom high-temperature gas-cooled reactor (HTGR) for 645 equivalent full power days. Four fuel varieties, contained in H-327 graphite bodies, were tested. A primary result of this test has been to demonstrate acceptable performance even with calculated high stresses in the graphite bodies. Heterogeneous fuel loadings in the element caused local power peaking and azimuthal power variations, deforming the graphite fuel bodies and thereby causing bowing nearly five times as large as the diametral clearance within the sleeve. The axial stresses resulting from interference between the fuel bodies and sleeve were estimated to have reached 45% of the ultimate material strength at the end of the irradiation. Residual stresses from differential contraction within the fuel body resulted in probable in-plane stress levels of 130% of the material strength at the end-of-life shutdown and of up to 150% of the strength at shutdown during the irradiation cycle. The high in-plane stresses are local peaks at the corners of a sharp notch in the element, which may account for the stresses failing to cause damage. The lack of observable damage, however, indicates that the methods and data used for stress analysis give results that are either fairly accurate or conservative

Prediction of the fuel failure following a large LOCA using modified gap heat transfer model

International Nuclear Information System (INIS)

Lee, K.M.; Lee, N.H.; Huh, J.Y.; Seo, S.K.; Choi, J.H.

1995-01-01

The modified Ross and Stoute gap heat transfer model in the ELOCA.Mk5 code for CANDU safety analysis is based on a simplified thermal deformation model. A review on a series of recent experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this study, more realistic offset crap conductance model is implemented in the code to estimate the fuel failure thresholds usincr the transient conditions of a 100% Reactor Outlet Header (ROH) break LOCA. Based on the offset gap conductance model, the total release of I-131 from the failed fuel elements in the core is reduced from 3876 TBq to 3283 TBq to increase margin for dose limit. (author)

System failure, innovation policy and patents: Fuel cells and related hydrogen technology in Norway 1990-2002

International Nuclear Information System (INIS)

Godoe, Helge; Nygaard, Stian

2006-01-01

The empirical focus of this article is technological innovation activities in the emerging field of fuel cells and related hydrogen technology in Norway from 1990 to 2002. In this period, four comparatively large-scale research and development projects and a number of smaller projects aimed at development of fuel cells technology were undertaken, resulting in many inventions that were subsequently patented. Although this creativity may be considered an indication of success, only one of the projects became successful in an innovation perspective. All the large projects were initiated and funded for divergent political and economic reasons. An important reason in the late 1980s was the prospect of using Norway's abundant supply of natural gas in fuel cells for electric power generation. The large R and D projects that attempted to develop fuel cells based on natural gas as energy source failed. In contrast, the successful project was undertaken by military R and D, i.e. in a different system of innovation than the projects that failed. Analysis of these cases points to the importance of a systemic approach to innovations-and to policy making. One challenge for policy makers is to decide how they should promote this development which is crucial for the vision of a future 'Hydrogen Economy', i.e. what kind of policy incentives should be introduced to spur efficiency in technological development and diffusion. Theoretically, many options are available; however, understanding the innovation dynamics in this sector is fundamental for making choices. In this article, focus will be set on policy aspects using an innovation systemic approach to analyze development of fuel cells and related hydrogen technology in Norway

Do helmets worn for hurling fail to protect the ear? Identification of an emerging injury pattern.

LENUS (Irish Health Repository)

Martin-Smith, James D

2012-12-01

Hurling is an Irish national game of stick and ball known for its ferocity, played by 190 000 players. Facial injuries were common but have been significantly reduced by legislation enforcing compulsory helmet wearing. Current standard helmets worn by hurlers do not offer protection to the external ear. Here we describe an emerging pattern of ear injuries and demonstrate the risk of external ear injuries in hurlers complying with current helmet safety standards. A 6-month retrospective analysis was carried out of patients attending Cork University Hospital (CUH) with ear lacerations sustained while hurling. Patient notes were reviewed and helmet manufacturers were interviewed. Seven patients were identified, all of whom sustained complex through ear lacerations while wearing helmets complying with current safety standards. Current helmet design fails to protect the external ear placing it at an increased risk of injury, a potential solution is to include ear protection in the helmet design.

Why did the League of Nations fail?

OpenAIRE

Jari Eloranta

2011-01-01

Why did the League of Nations ultimately fail to achieve widespread disarmament, its most fundamental goal? This article shows that the failure of the League of Nations had two important dimensions: (1) the failure to provide adequate security guarantees for its members (like an alliance); (2) the failure of this organization to achieve the disarmament goals it set out in the 1920s and 1930s. Thus, it was doomed from the outset to fail, due to built-in institutional contradictions. It can als...

UO2 Fuel pellet impurities, pellet surface roughness and n(18O)/n(16O) ratios, applied to nuclear forensic science

International Nuclear Information System (INIS)

Pajo, L.

2001-01-01

In the last decade, law enforcement has faced the problem of illicit trafficking of nuclear materials. Nuclear forensic science is a new branch of science that enables the identification of seized nuclear material. The identification is not based on a fixed scheme, but further identification parameters are decided based on previous identification results. The analysis is carried out by using traditional analysis methods and applying modern measurement technology. The parameters are generally not unambiguous and not self-explanatory. In order to have a full picture about the origin of seized samples, several identification parameters should be used together and the measured data should be compared to corresponding data from known sources. A nuclear material database containing data from several fabrication plants is installed for the purpose. In this thesis the use of UO 2 fabrication plant specific parameters, fuel impurities, fuel pellet surface roughness and oxygen isotopic ratio in UO 2 were investigated for identification purposes in nuclear forensic science. The potential use of these parameters as 'fingerprints' is discussed for identification purposes of seized nuclear materials. Impurities of the fuel material vary slightly according to the fabrication method employed and a plant environment. Here the impurities of the seized UO 2 were used in order to have some clues about the origin of the fuel material by comparing a measured data to nuclear database information. More certainty in the identification was gained by surface roughness of the UO 2 fuel pellets, measured by mechanical surface profilometry. Categories in surface roughness between a different fuel element type and a producer were observed. For the time oxygen isotopic ratios were determined by Thermal Ionisation Mass Speckometry (TIMS). Thus a TIMS measurement method, using U 16 O + and U 18 0 + ions, was developed and optimised to achieve precise oxygen isotope ratio measurements for the

Serus, an expert system for the ultrasonic examination of fuel rods

International Nuclear Information System (INIS)

Gondard, C.; Papezyk, F.; Wident, P.

1987-01-01

The use of pattern recognition functions and the modelization of the human expert reasoning, allow the automatic identification of defects in welds or structures. The proposed application uses an ultrasonic examination to detect and classify 3 types of defects in end plug welds of PWR fuel rods

Calculation of the fuel temperature field under heat release and heat conductance transient conditions

International Nuclear Information System (INIS)

Kazakov, E.K.; Chernukhina, G.M.

1974-01-01

Results of calculation of the temperature distribution in an annular fuel element at transient thermal conductivity and heat release values are given. The calculation has been carried out by the mesh technique with the third-order boundary conditions for the inner surface assumed and with heat fluxes and temperatures at the zone boundaries to be equal. Three variants of solving the problem of a stationary temperature field are considered for failed fuel elements with clad flaking or cracks. The results obtained show the nonuniformity of the fuel element temperature field to depend strongly on the perturbation parameter at transient thermal conductivity and heat release values. In case of can flaking at a short length, the core temperature rises quickly after flaking. While evaluating superheating, one should take into account the symmetry of can flaking [ru

Relative Release-to-Birth Indicators for Investigating TRISO Fuel Fission Gas Release Models

International Nuclear Information System (INIS)

Harp, Jason M.; Hawari, Ayman I.

2008-01-01

TRISO microsphere fuel is the fundamental fuel unit for Very High Temperature Reactors (VHTR). A single TRISO particle consists of an inner kernel of uranium dioxide or uranium oxycarbide surrounded by layers of pyrolytic carbon and silicon carbide. If the silicon carbide layer fails, fission products, especially the noble fission gases Kr and Xe, will begin to escape the failed particle. The release of fission gas is usually quantified by measuring the ratio of the released activity (R) to the original birth activity (B), which is designated as the R/B ratio. In this work, relative Release-to-Birth indicators (I) are proposed as a technique for interpreting the results of TRISO irradiation experiments. By implementing a relative metric, it is possible to reduce the sensitivity of the indicators to instrumental uncertainties and variations in experimental conditions. As an example, relative R/B indicators are applied to the interpretation of representative data from the Advanced Gas Reactor-1 TRISO fuel experiment that is currently taking place at the Advanced Test Reactor of Idaho National Laboratory. It is shown that the comparison of measured to predicted relative R/B indicators (I) gives insight into the physics of release and helps validate release models. Different trends displayed by the indicators are related to the mechanisms of fission gas release such as diffusion and recoil. The current analysis shows evidence for separate diffusion coefficients for Kr and Xe and supports the need to account for recoil release. (authors)

Developing RCM Strategy for Hydrogen Fuel Cells Utilizing On Line E-Condition Monitoring

International Nuclear Information System (INIS)

Baglee, D; Knowles, M J

2012-01-01

Fuel cell vehicles are considered to be a viable solution to problems such as carbon emissions and fuel shortages for road transport. Proton Exchange Membrane (PEM) Fuel Cells are mainly used in this purpose because they can run at low temperatures and have a simple structure. Yet high maintenance costs and the inherent dangers of maintaining equipment using hydrogen are two main issues which need to be addressed. The development of appropriate and efficient strategies is currently lacking with regard to fuel cell maintenance. A Reliability Centered Maintenance (RCM) approach offers considerable benefit to the management of fuel cell maintenance since it includes an identification and consideration of the impact of critical components. Technological developments in e-maintenance systems, radio-frequency identification (RFID) and personal digital assistants (PDAs) have proven to satisfy the increasing demand for improved reliability, efficiency and safety. RFID technology is used to store and remotely retrieve electronic maintenance data in order to provide instant access to up-to-date, accurate and detailed information. The aim is to support fuel cell maintenance decisions by developing and applying a blend of leading-edge communications and sensor technology including RFID. The purpose of this paper is to review and present the state of the art in fuel cell condition monitoring and maintenance utilizing RCM and RFID technologies. Using an RCM analysis critical components and fault modes are identified. RFID tags are used to store the critical information, possible faults and their cause and effect. The relationship between causes, faults, symptoms and long term implications of fault conditions are summarized. Finally conclusions are drawn regarding suggested maintenance strategies and the optimal structure for an integrated, cost effective condition monitoring and maintenance management system.

Developing RCM Strategy for Hydrogen Fuel Cells Utilizing On Line E-Condition Monitoring

Science.gov (United States)

Baglee, D.; Knowles, M. J.

2012-05-01

Fuel cell vehicles are considered to be a viable solution to problems such as carbon emissions and fuel shortages for road transport. Proton Exchange Membrane (PEM) Fuel Cells are mainly used in this purpose because they can run at low temperatures and have a simple structure. Yet high maintenance costs and the inherent dangers of maintaining equipment using hydrogen are two main issues which need to be addressed. The development of appropriate and efficient strategies is currently lacking with regard to fuel cell maintenance. A Reliability Centered Maintenance (RCM) approach offers considerable benefit to the management of fuel cell maintenance since it includes an identification and consideration of the impact of critical components. Technological developments in e-maintenance systems, radio-frequency identification (RFID) and personal digital assistants (PDAs) have proven to satisfy the increasing demand for improved reliability, efficiency and safety. RFID technology is used to store and remotely retrieve electronic maintenance data in order to provide instant access to up-to-date, accurate and detailed information. The aim is to support fuel cell maintenance decisions by developing and applying a blend of leading-edge communications and sensor technology including RFID. The purpose of this paper is to review and present the state of the art in fuel cell condition monitoring and maintenance utilizing RCM and RFID technologies. Using an RCM analysis critical components and fault modes are identified. RFID tags are used to store the critical information, possible faults and their cause and effect. The relationship between causes, faults, symptoms and long term implications of fault conditions are summarized. Finally conclusions are drawn regarding suggested maintenance strategies and the optimal structure for an integrated, cost effective condition monitoring and maintenance management system.

Failed total carpometacarpal joint prosthesis of the thumb

DEFF Research Database (Denmark)

Hansen, Torben Bæk; Homilius, Morten

2010-01-01

Total joint prosthesis in carpometacarpal joint arthritis of the thumb often fails. Loosening of the implant is often treated by resection arthroplasty, and we reviewed 10 patients, mean age 54 years (range 47-63) who were treated by resection arthroplasty after a failed total joint prosthesis. T...... in eight of 10 patients, but the mean Disabilities of the arm, shoulder, and hand (DASH) scores, self-reported pinch-grip-related function, and pain were comparable with our earlier published results with the Elektra carpometacarpal total joint prosthesis.......Total joint prosthesis in carpometacarpal joint arthritis of the thumb often fails. Loosening of the implant is often treated by resection arthroplasty, and we reviewed 10 patients, mean age 54 years (range 47-63) who were treated by resection arthroplasty after a failed total joint prosthesis....... The male:female ratio was 1:4 and the mean duration of observation 32 months (range 6-52). In three patients the revised implant was a MOJE uncemented carpometacarpal joint prosthesis and in seven patients an Elektra uncemented one. At follow-up grip strength was reduced to less than 90% of the other hand...

German Approach to Spent Fuel Management

International Nuclear Information System (INIS)

Jussofie, A.; Graf, R.; Filbert, W.

2010-01-01

The management of spent fuel was based on two powerful columns until 30 June 2005, i. e. reprocessing and direct disposal. After this date any delivery of spent fuel to reprocessing plants was prohibited so that the direct disposal of unreprocessed spent fuel is the only available option in Germany today. The main steps of the current concept are: (i) Intermediate storage of spent fuel, which is the only step in practice. After the first cooling period in spent fuel storage pools it continues into cask-receiving dry storage facilities. Identification of casks, 'freezing' of inventories in terms of continuity of knowledge, monitoring the access to spent fuel, verifying nuclear material movements in terms of cask transfers and ensurance against diversion of nuclear material belong to the fundamental safeguards goals which have been achieved in the intermediate storage facilities by containment and surveillance techniques in unattended mode. (ii) Conditioning of spent fuel assemblies by separating the fuel rods from structural elements. Since the pilot conditioning facility in Gorleben has not yet come into operation, the underlying safeguards approach which focuses on safeguarding the key measurement points - the spent fuel related way in and out of the facility - has not been applied yet. (iii) Disposal in deep geological formations, but no decision has been made so far neither regarding the location of a geological repository nor regarding the safeguards approach for the disposal concept of spent fuel. The situation was complicated by a moratorium which suspended the underground exploration of the Gorleben salt dome as potential geological repository for spent fuel. The moratorium expires in October 2010. Nevertheless, considerable progress has been made in the development of disposal concepts. According to the basic, so-called POLLUX (registered) -concept spent fuel assemblies are to be conditioned after dry storage and reloaded into the POLLUX (registered) -cask

Identification and Reconfigurable Control of Impaired Multi-Rotor Drones

Science.gov (United States)

Stepanyan, Vahram; Krishnakumar, Kalmanje; Bencomo, Alfredo

2016-01-01

The paper presents an algorithm for control and safe landing of impaired multi-rotor drones when one or more motors fail simultaneously or in any sequence. It includes three main components: an identification block, a reconfigurable control block, and a decisions making block. The identification block monitors each motor load characteristics and the current drawn, based on which the failures are detected. The control block generates the required total thrust and three axis torques for the altitude, horizontal position and/or orientation control of the drone based on the time scale separation and nonlinear dynamic inversion. The horizontal displacement is controlled by modulating the roll and pitch angles. The decision making algorithm maps the total thrust and three torques into the individual motor thrusts based on the information provided by the identification block. The drone continues the mission execution as long as the number of functioning motors provide controllability of it. Otherwise, the controller is switched to the safe mode, which gives up the yaw control, commands a safe landing spot and descent rate while maintaining the horizontal attitude.

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

International Nuclear Information System (INIS)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Energy Technology Data Exchange (ETDEWEB)

Bond, W.D.; Mailen, J.C.; Michaels, G.E.

1992-10-01

The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

1993-01-01

Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

Comparison of Tagging Technologies for Safeguards of Copper Canisters for Nuclear Spent Fuel.

Science.gov (United States)

Clementi, Chiara; Littmann, François; Capineri, Lorenzo

2018-03-21

Several countries are planning to store nuclear spent fuel in long term geological repositories, preserved by copper canisters with an iron insert. This new approach involves many challenging problems and one is to satisfy safeguards requirements: the Continuity of Knowledge (CoK) of the fuel must be kept from the encapsulation plant up to the final repository. To date, no measurement system has been suggested for a unique identification and authentication. Following the list of the most important safeguards, safety and security requirements for copper canisters identification and authentication, a review of conventional tagging technologies and measurement systems for nuclear items is reported in this paper. The aim of this study is to verify to what extent each technology could be potentially used for keeping the CoK of copper canisters. Several tagging methods are briefly described and compared, discussing advantages and disadvantages.

Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

International Nuclear Information System (INIS)

Monteiro, Iara Arraes

1999-02-01

The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

Thermal analysis of the failed equipment storage vault system

International Nuclear Information System (INIS)

Jerrell, J.; Lee, S.Y.; Shadday, A.

1995-07-01

A storage facility for failed glass melters is required for radioactive operation of the Defense Waste Processing Facility (DWPF). It is currently proposed that the failed melters be stored in the Failed Equipment Storage Vaults (FESV's) in S area. The FESV's are underground reinforced concrete structures constructed in pairs, with adjacent vaults sharing a common wall. A failed melter is to be placed in a steel Melter Storage Box (MSB), sealed, and lowered into the vault. A concrete lid is then placed over the top of the FESV. Two melters will be placed within the FESV/MSB system, separated by the common wall. There is no forced ventilation within the vault so that the melter is passively cooled. Temperature profiles in the Failed Equipment Storage Vault Structures have been generated using the FLOW3D software to model heat conduction and convection within the FESV/MSB system. Due to complexities in modeling radiation with FLOW3D, P/THERMAL software has been used to model radiation using the conduction/convection temperature results from FLOW3D. The final conjugate model includes heat transfer by conduction, convection, and radiation to predict steady-state temperatures. Also, the FLOW3D software has been validated as required by the technical task request

Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

Energy Technology Data Exchange (ETDEWEB)

Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2016-04-01

In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

Plan for spent fuel waste form testing for NNWSI [Nevada Nuclear Waste Storage Investigations

International Nuclear Information System (INIS)

Shaw, H.F.

1987-11-01

The purpose of spent fuel waste form testing is to determine the rate of release of radionuclides from failed disposal containers holding spent fuel, under conditions appropriate to the Nevada Nuclear Waste Storage Investigations (NNWSI) Project tuff repository. The information gathered in the activities discussed in this document will be used: to assess the performance of the waste package and engineered barrier system (EBS) with respect to the containment and release rate requirements of the Nuclear Regulatory Commission, as the basis for the spent fuel waste form source term in repository-scale performance assessment modeling to calculate the cumulative releases to the accessible environment over 10,000 years to determine compliance with the Environmental Protection Agency, and as the basis for the spent fuel waste form source term in repository-scale performance assessment modeling to calculate cumulative releases over 100,000 years as required by the site evaluation process specified in the DOE siting guidelines. 34 refs

Performance evaluation of UO2-Zr fuel in power ramp tests

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.

1977-01-01

In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

Identification of fast power reactivity effect in nuclear power reactor

International Nuclear Information System (INIS)

Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

1987-01-01

A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1993-01-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with an analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (1) intergranular stress-corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor

Analysis of failed nuclear plant components

International Nuclear Information System (INIS)

Diercks, D.R.

1992-07-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (a) intergranular stress corrosion cracking of core spray injection piping in a boiling water reactor, (b) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressure water reactor, (c) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (d) failure of pump seal wear rings by nickel leaching in a boiling water reactor

High Ethanol Fuel Endurance: A Study of the Effects of Running Gasoline with 15% Ethanol Concentration in Current Production Outboard Four-Stroke Engines and Conventional Two-Stroke Outboard Marine Engines

Energy Technology Data Exchange (ETDEWEB)

Hilbert, D.

2011-10-01

Three Mercury Marine outboard marine engines were evaluated for durability using E15 fuel -- gasoline blended with 15% ethanol. Direct comparison was made to operation on E0 (ethanol-free gasoline) to determine the effects of increased ethanol on engine durability. Testing was conducted using a 300-hour wide-open throttle (WOT) test protocol, a typical durability cycle used by the outboard marine industry. Use of E15 resulted in reduced CO emissions, as expected for open-loop, non-feedback control engines. HC emissions effects were variable. Exhaust gas and engine operating temperatures increased as a consequence of leaner operation. Each E15 test engine exhibited some deterioration that may have been related to the test fuel. The 9.9 HP, four-stroke E15 engine exhibited variable hydrocarbon emissions at 300 hours -- an indication of lean misfire. The 300HP, four-stroke, supercharged Verado engine and the 200HP, two-stroke legacy engine tested with E15 fuel failed to complete the durability test. The Verado engine failed three exhaust valves at 285 endurance hours while the 200HP legacy engine failed a main crank bearing at 256 endurance hours. All E0-dedicated engines completed the durability cycle without incident. Additional testing is necessary to link the observed engine failures to ethanol in the test fuel.

Source identification of underground fuel spills by solid-phase microextraction/high-resolution gas chromatography/genetic algorithms.

Science.gov (United States)

Lavine, B K; Ritter, J; Moores, A J; Wilson, M; Faruque, A; Mayfield, H T

2000-01-15

Solid-phase microextraction (SPME), capillary column gas chromatography, and pattern recognition methods were used to develop a potential method for typing jet fuels so a spill sample in the environment can be traced to its source. The test data consisted of gas chromatograms from 180 neat jet fuel samples representing common aviation turbine fuels found in the United States (JP-4, Jet-A, JP-7, JPTS, JP-5, JP-8). SPME sampling of the fuel's headspace afforded well-resolved reproducible profiles, which were standardized using special peak-matching software. The peak-matching procedure yielded 84 standardized retention time windows, though not all peaks were present in all gas chromatograms. A genetic algorithm (GA) was employed to identify features (in the standardized chromatograms of the neat jet fuels) suitable for pattern recognition analysis. The GA selected peaks, whose two largest principal components showed clustering of the chromatograms on the basis of fuel type. The principal component analysis routine in the fitness function of the GA acted as an information filter, significantly reducing the size of the search space, since it restricted the search to feature subsets whose variance is primarily about differences between the various fuel types in the training set. In addition, the GA focused on those classes and/or samples that were difficult to classify as it trained using a form of boosting. Samples that consistently classify correctly were not as heavily weighted as samples that were difficult to classify. Over time, the GA learned its optimal parameters in a manner similar to a perceptron. The pattern recognition GA integrated aspects of strong and weak learning to yield a "smart" one-pass procedure for feature selection.

Failing States or Failing Models?: Accounting for the Incidence of State Collapse

OpenAIRE

Martin Doornbos

2010-01-01

In recent years the notion and phenomenon of .failingÿ states - states deemed incapable to fulfil the basic tasks of providing security for their populace -, has been rapidly drawing attention. I will start off with a closer look at the inci- dence of fragile states and state failure, more specifically of state collapse. Connected with this, I will raise the question of differential degrees of propensity to failure and collapse among contemporary state systems, and to point to apparent region...

The dynamics of female self-identification conceptions: from modernism to postmodernist feminism

OpenAIRE

Е Е Agafonova

2009-01-01

A number of new contradictions between the personality and society, including the accentuation of the problem of human being's identity and self-identification have been engendered by globalization. Whereas formerly the individual enjoyed a stable set of norms, values, and motivations determining his identity structure, nowadays social institutions bearing responsibility for the stable identity development are experiencing transformations and fail to accomplish their functions to the full ext...

Odor identification: perceptual and semantic dimensions.

Science.gov (United States)

Cain, W S; de Wijk, R; Lulejian, C; Schiet, F; See, L C

1998-06-01

Five studies explored identification of odors as an aspect of semantic memory. All dealt in one way or another with the accessibility of acquired olfactory information. The first study examined stability and showed that, consistent with personal reports, people can fail to identify an odor one day yet succeed another. Failure turned more commonly to success than vice versa, and once success occurred it tended to recur. Confidence ratings implied that subjects generally knew the quality of their answers. Even incorrect names, though, often carried considerable information which sometimes reflected a semantic and sometimes a perceptual source of errors. The second study showed that profiling odors via the American Society of Testing and Materials list of attributes, an exercise in depth of processing, effected no increment in the identifiability/accessibility beyond an unelaborated second attempt at retrieval. The third study showed that subjects had only a weak ability to predict the relative recognizability of odors they had failed to identify. Whereas the strength of the feeling that they would 'know' an answer if offered choices did not associate significantly with performance for odors, it did for trivia questions. The fourth study demonstrated an association between ability to discriminate among one set of odors and to identify another, but this emerged only after subjects had received feedback about identity, which essentially changed the task to one of recognition and effectively stabilized access. The fifth study illustrated that feedback improves performance dramatically only for odors involved with it, but that mere retrieval leads to some improvement. The studies suggest a research agenda that could include supplemental use of confidence judgments both retrospectively and prospectively in the same subjects to indicate the amount of accessible semantic information; use of second and third guesses to examine subjects' simultaneously held hypotheses about

About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

International Nuclear Information System (INIS)

Grabezhnaya, V.A.

2008-01-01

Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

Radionuclide release from PWR fuels in a reference tuff repository groundwater

International Nuclear Information System (INIS)

Wilson, C.N.; Oversby, V.M.

1985-03-01

The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-μm wide slit through the cladding; rod sections with water-tight end fittings and two 200-μm-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO 2 spent fuel matrix dissolves. Fractional release of 137 Cs and 99 Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables

The microstructure of fuel pellets as object of quality characterization on base of FMEA analysis

International Nuclear Information System (INIS)

Goncharov, U.V.; Matveev, A.A.; Strucov, A.V.; Loktev, I.I.

2012-01-01

It is difficult to find new effective reserves in nuclear fuel production as its experience of production and operation become more and more. FMEA method can help it on base of the system analysis. The state corporation Rosatom, consistently pursuing a policy of economical manufacture, make all efforts for identification of deep dependences between conditions of manufacture, characteristics of fuel materials and features of their operational behaviour. This report continues earlier discussion of the important feature of produced nuclear fuel pellets grain size distribution. This distribution defines gas release in reactor and has not appropriate method of characterization. There are descriptions of optimal microstructure of fuel pellets with large grain size literature

Preliminary Identification of the Microbial Fuel Cell Bacteria Communities in Sewage

International Nuclear Information System (INIS)

Zain, S.M; Hashim, R.; Roslani, N.S.; Suja, F.; Basri, N.E.A.; Anuar, N.; Daud, W.R.W.

2011-01-01

This study aimed to determine the types of bacteria exist in wastewater that contribute to generate electricity and simultaneously remove carbon and nitrogen. The method used was Fluorescence In Situ Hybridization (FISH) to detect the bacteria group while Polymerase Chain Reaction (PCR) was used to confirm the observation made using FISH. A biochemical identification using BIOLOG GEN III MICROPLATE TM also was used . The samples were cultured on nutrient agar plate to identify the morphology of the bacteria. The result showed that 21 isolates from three different locations at the activated sludge treatment plant with six, eleven and four strains for raw sewage, aeration tank and returned activated sludge samples, respectively. Preliminary identification does not give a good match but only showed the existence of ammonia-oxidizing bacteria (FISH) and Kurtia Gibsoni (BIOLOG) from aeration tank : Bacillus sp (PCR) and Bacillus Pseudomycoides (BIOLOG) from returned activated sludge. The maximum power density generated using returned activated sludge was 9.053 mW/ cm 2 , with 26.8 % COD removal and 40 % TKN removal (author)

Analysis of failed nuclear plant components

Science.gov (United States)

Diercks, D. R.

1993-12-01

Argonne National Laboratory has conducted analyses of failed components from nuclear power- gener-ating stations since 1974. The considerations involved in working with and analyzing radioactive compo-nents are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in serv-ice. The failures discussed are (1) intergranular stress- corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

Solid oxide fuel cell anode image segmentation based on a novel quantum-inspired fuzzy clustering

Science.gov (United States)

Fu, Xiaowei; Xiang, Yuhan; Chen, Li; Xu, Xin; Li, Xi

2015-12-01

High quality microstructure modeling can optimize the design of fuel cells. For three-phase accurate identification of Solid Oxide Fuel Cell (SOFC) microstructure, this paper proposes a novel image segmentation method on YSZ/Ni anode Optical Microscopic (OM) images. According to Quantum Signal Processing (QSP), the proposed approach exploits a quantum-inspired adaptive fuzziness factor to adaptively estimate the energy function in the fuzzy system based on Markov Random Filed (MRF). Before defuzzification, a quantum-inspired probability distribution based on distance and gray correction is proposed, which can adaptively adjust the inaccurate probability estimation of uncertain points caused by noises and edge points. In this study, the proposed method improves accuracy and effectiveness of three-phase identification on the micro-investigation. It provides firm foundation to investigate the microstructural evolution and its related properties.

Hardware in the loop simulation test platform of fuel cell backup system

Directory of Open Access Journals (Sweden)

Ma Tiancai